assessment of the neutron and gamma sources of the spent

Aapo TanskanenVTT Energy

In STUK this study was supervised by Matti Tarvainen

Assessment of the neutronand gamma sources of thespent BWR fuel

Interim report on Task FIN JNT A 1071 of

the Finnish support programme to IAEA

Safeguards

October 2000S T U K - Y T O - T R 1 7 0

S T U K • S Ä T E I L Y T U R V A K E S K U S • S T R Å L S Ä K E R H E T S C E N T R A L E NR A D I A T I O N A N D N U C L E A R S A F E T Y A U T H O R I T Y

ISBN 951-712-419-8ISSN 0785-9325

Oy Edita Ab, Helsinki 2000

The conclusions presented in the STUK report series are those of the authorsand do not necessarily represent the official position of STUK.

S T U K - Y T O - T R 1 7 0

3

The neutron and gamma sources of the spent nuclear fuel were assessed by calculating the nuclideinventory of the spent fuel using SAS2H/ORIGEN-S code. The calculations were performed by VTTEnergy under Task JNT A 1071 FIN Partial Defect Test on Spent Fuel LWRs. According to thecalculation results the neutron emission of the spent LWR fuel is dominantly from spontaneous fissionof Cm-244 at cooling times longer than two years. The parameters that affect the Cm-244 concentrationat discharge are fuel bundle type, burnup, and in BWR also void fraction. The power history has only aslight influence on buildup of Cm-244. Also the axial distribution of the neutron source can be deter-mined. Spent LWR fuel contains a variety of gamma emitting nuclides. However, only Cs-137 and Cs-134 gamma peaks stand out from the gamma spectrum. Cs-137 is a better indicator of the burnup thanCs-134, but the ratio between the amount of Cs-134 and Cs-137 can be used to estimate the coolingtime.

ABSTRACT

TANSKANEN Aapo (VTT Energy). Assessment of the neutron and gamma sources of the spent BWRfuel. Interim report on Task FIN JNT A 1071 of the Finnish Support Programme to IAEA Safeguards.STUK-YTO-TR 170. Helsinki 2000. 17 pp + Appendices 14 pp.

ISBN 951-712-419-8ISSN 0785-9325

Keywords: nuclear reactors, light water reactors, spent fuels, burnup, neutron sources,gamma sources

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S T U K - Y T O - T R 1 7 0

CONTENTS

ABSTRACT 3

CONTENTS 4

1 INTRODUCTION 5

2 CALCULATIONAL TOOLS 62.1 SAS2H/ORIGEN-S in SCALE 62.2 CASMO-4 6

3 SPENT LWR FUEL NEUTRON AND GAMMA SOURCES 73.1 Nuclide inventory calculations 73.1 The major components of the spent LWR fuel neutron and gamma sources 73.3 The accuracy of the calculational results 8

4 NEUTRON SOURCE ASSESSMENT 94.1 Cooling time decay 94.2 Simple 244Cm neutron source model 94.3 Enrichment dependence 104.4 Void fraction dependence 114.5 Effect of the amount of Gd2O3 in burnable absorber rods 124.6 Axial neutron source distribution 124.7 Pinwise neutron source distribution 134.8 Assessment of the neutron source of the Olkiluoto BWR bundles 13

5 ASSESSMENT OF THE 137Cs AND 134Cs GAMMA SOURCES 15

6 CONCLUSIONS 16

REFERENCES 17

APPENDIX A Nuclide inventory of spent BWR fuel 18APPENDIX B Nuclide inventory of spent VVER-440 fuel 21APPENDIX C Neutron source of spent BWR fuel 25APPENDIX D Gamma source of spent BWR fuel 29

S T U K - Y T O - T R 1 7 0

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1 INTRODUCTION

LWR spent fuel assemblies are being placed instorage facilities where they are considered diffi-cult to access. According to the safeguards con-cept, IAEA verifies the assemblies before they be-come difficult to access. Therefore, a standard ver-ification procedure needs to be developed. The aimof this work was to calculationally assess the neu-tron and gamma source of the spent LWR fuel.The calculations were performed by VTT Energyunder Task JNT A 1071 FIN Partial Defect Test onSpent Fuel LWRs.

In this work the nuclide inventory of the spentBWR and VVER-440 type fuels were calculatedapplying SAS2H/ORIGEN-S code that is part ofthe modular SCALE code system [1]. CASMO-4was applied to the BWR case as an additionalreference code. In parallel with this work, theneutron emission rate and gamma spectrum ofsome BWR bundles of the Olkiluoto reactors weremeasured with IAEA BWR Fork detector coupledwith compact underwater gamma spectrometer[2].

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S T U K - Y T O - T R 1 7 0

2.1 SAS2H/ORIGEN-S in SCALE

The SCALE (Standardized Computer Analysis forLicencing Evaluation) system was developed byOak Ridge National Laboratory for the NuclearRegulatory Commission to satisfy a need for astandardized method of analysis for the evalua-tion of nuclear fuel facility and package designs.In its present form, the system has the capabilityto perform criticality, shielding, and heat transferanalyses using well established functional mod-ules tailored to the SCALE system. In this workSCALE-4.4 acquired through NEA Data Bank wasapplied for assessment of the neutron source ofspent LWR fuel.

ORIGEN-S is a functional module of theSCALE system that performs a depletion anddecay calculation for spent fuel characterizationand source term generation. A matrix exponentialexpansion model supplemented with Batemannchain equations is used to solve the nuclide gener-ation and depletion equations. Over 1600 nuclidesare followed in the depletion analysis [1, S2.2.10].The SAS2H module uses ORIGEN-S to perform afuel depletion analysis (to characterize spent fueland/or generate source terms) and an optionalone-dimensional radial shielding analysis of acask-type geometry. SAS2H execution includes:repeated passes through BONAMI, NITAWL-II,XSDRNPM, COUPLE, and ORIGEN-S. This sim-ple procedure has been shown to produce conserv-

ative actinide inventories for both PWR and BWRspent fuel. The neutron flux used to update thecross sections during the depletion was based on aone-dimensional model of the fuel assembly. Eventhough the flux in the axial direction of theassembly is assumed to be constant, the procedurecharacterizes the major 2-D effects.

Several multigroup cross section data librariesare available for SAS2H/ORIGEN-S depletionanalysis. In this work 44-group 44GROUPNDF5cross section library was used. It is based on theENDF/B-V evaluated nuclear data. SAS2H/ORI-GEN-S has been validated against several west-ern type PWRs and BWRs [1, S2.3.1]. Validationstudies for BWR have been conducted but they arestill to be published.

2.2 CASMO-4

CASMO-4 is a multigroup, two-dimensional trans-port theory code for burnup calculations on BWRand PWR fuel assemblies. The code can handlegeometries consisting of cylindrical fuel rods ofvarying composition in a square or hexagonal ar-ray. Two multigroup cross section data librariesare available for CASMO-4 calculations: 70-groupE4LBL70 and a condensed 40-group E4LBL40.Both libraries are based mainly on the ENDF/B-IV evaluated nuclear data. In this work CASMO-4was applied as an additional reference code.

2 CALCULATIONAL TOOLS

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7

3.1 Nuclide inventory calculations

In this work SAS2H/ORIGEN-S and CASMO-4were applied to assess the nuclide inventory ofspent BWR fuel. The BWR fuel bundle was a com-mon 8x8 BWR fuel bundle with four gadoliniumburnable absorber rods. The amount of Gd2O3 inthe burnable absorber rods was 2 w-%. The fuelaverage enrichment was 2.75 w-%, its average dis-charge burnup was 37.8 MWd/kgU, and the aver-age coolant density was 0.426 g/cm3 which corre-sponds to average void fraction 0.45. The resultsof the SAS2H/ORIGEN-S calculations in AppendixA give the nuclide inventory right after discharge(column: initial) and at several cooling times.SAS2H/ORIGEN-S was also applied to assess thenuclide inventory of spent VVER-440 fuel. TheVVER-440 assembly studied was a typical uni-formly enriched (3,6 w-%) hexagonal bundle with126 fuel rods and a central water rod. Its averagedischarge burnup was assumed to be 36 MWd/kgU. The results of the SAS2H/ORIGEN-S calcu-lation are in Appendix B.

3.1 The major components of thespent LWR fuel neutron andgamma sources

In appendices C and D are the major componentsof the primary neutron and gamma sources of theabove described spent BWR fuel bundles calculat-ed by SAS2H/ORIGEN-S. The major componentsof the neutron and gamma sources of the spentVVER-440 fuel were basically the same. All thesources were normalized to an assembly. ORI-GEN-S gives the gamma source spectrum in eight-een gamma energy groups so that the sourceterms are normalized to the average energy ofeach group in order to preserve the total gammaenergy production.

The major components of the calculated neu-tron source at discharge are shown in Table I andTable II. According to the calculational results themajor part of the neutron source is produced fromspontaneous fission of heavy nuclides, namely244Cm and 242Cm. A significant neutron source isalso produced from 17O(a,n) and 18O(a,n) reactionsin the UO2 and other oxygen compounds of thespent fuel.

Spent LWR fuel contains a variety of gammaemitting nuclides. After ten years cooling time137Cs or actually its daughter nuclide 137mBa is thedominant gamma emitter. It has been shown thatthe amount of 137Cs depends linearly on burnup[3], and with gamma spectroscopy the 137Cs con-tents of spent fuel can be determined based on themeasured 662 keV gamma line emission. Othermajor gamma emitters are 154Eu, 134Cs, 106Ru and

3 SPENT LWR FUEL NEUTRON AND GAMMASOURCES

Table I. The major components of the neutron sourceof BWR spent fuel at discharge (n/s/assembly),burnup was 37.8 MWd/kgU.

Table II. The major components of the neutronsource of VVER-440 spent fuel at discharge (n/s/assembly), burnup was 36 MWd/kgU.

stnenopmoC S-NEGIRO/H2SAS 4-OMSAC

fo.F.S 442 mC 01·24.1 8 01·79.9 7

fo.F.S 242 mC 01·89.6 7 01·46.6 7

latotnoissifsuoenatnopS 01·41.2 8 01·76.1 8

(a noitcaer-)n, 01·85.1 7 01·02.8 7

latoT 01·03.2 8 01·94.2 8

stnenopmoC S-NEGIRO/H2SAS

fo.F.S 442 mC 01·78.4 7

fo.F.S 242 mC 01·55.3 7

latotnoissifsuoenatnopS 01·05.8 7

(a noitcaer-)n, 01·98.7 6

latoT 01·92.9 7

8

S T U K - Y T O - T R 1 7 0

144Ce [4]. However, last two have a relatively shorthalf-life.

3.3 The accuracy of thecalculational results

The average difference in the SAS2H/ORIGEN-Scomputed and measured PWR isotopic concentra-tion for 244Cm has been found to be -9% whenapplying 44GROUPNDF5 nuclear data library [1,S2.3.1]. The combined uncertainty in the comput-ed neutron source from spontaneous fission is inthe range 4 to 8%, with a decrease as a function oftime resulting from the decay of 242Cm. In ORI-GEN-S thin target cross sections for (a,n)-reac-tions and alpha stopping powers data are appliedto compute neutron yields of the fuel material.The data and method originally used in ORIGENoverpredicted the (a,n) yields by a factor of ap-proximately 2. This is in accordance with the cal-culations of Anttila using ORIGEN-2 [5]. The un-certainty of the present data and method is esti-mated to be of the order of 20%. Thus, overalluncertainty of the total neutron source after 5

years cooling time is estimated to be of the orderof 10%.

According to the calculational results CASMO-4 seems to slightly underpredict the neutronsource from spontaneous fission, and drasticallyoverpredict neutrons from (a,n)-reactions. Never-theless, the total neutron source given by CAS-MO-4 agrees relatively well with the SAS2H/ORIGEN-S result. The cross section data used inthe CASMO-4 calculations was based on ENDF/B-IV evaluated nuclear data library, and the deple-tion calculation of CASMO-4 follows only some100 most important nuclides. The data and meth-od used to estimate (a,n) yields is basically similarto that of ORIGEN-2.

The accuracy of gamma source spectra is de-pendent upon the energy. Photon rates computedfor fission products tend to be more accurate thanthose for actinides, because the fission productinventory computed for a given burnup dependsdirectly upon calculation of removable energy perfission. Thus, the uncertainty of the 137Cs and134Cs gamma sources should be of the order of 5%[1, S2.3.3].

S T U K - Y T O - T R 1 7 0

9

0,0E+00

1,0E+08

2,0E+08

3,0E+08

0 5 10 15 20 25 30 35 40

Cooling time (a)

Neu

tro

n so

urce

(n/

s/as

sem

bly

)

Total

Cm-244

Cm-242

(a,n)

4.1 Cooling time decay

The decay of the major components of the neutronsource during cooling is shown in Figure 1. Atdischarge (a,n)-reactions contribute 7% of the neu-tron source, and their contribution decreases inthe course of time. 242Cm contributes about 30% ofthe neutron source at discharge, but due to itseshort half-life (163 d) it decays leaving 244Cm thedominant neutron source at longer cooling times.

After discharge 244Cm decays with a half-life18.1 years, and a straightforward correction canbe introduced to account for it

0( ) cTp c pS T S e l-

= (1),

where 0pS is the primary 244Cm neutron source at

discharge,l is the decay constant of the 244Cm

(lÿ= 1.2135×10-9 1/s), andTc is the cooling time after discharge.

The cooling time correction based on the decay of244Cm was verified by SAS2H/ORIGEN-S calcula-tion, where the neutron source from spontaneousfission of 244Cm was calculated as a function oftime. The cooling lasted for 50 years, and the re-sults given by a simple decay model agreed wellwith the SAS2H/ORIGEN-S results.

4.2 Simple 244Cm neutron sourcemodel

According to Würz [6] The assembly-averaged pri-mary 244Cm neutron source at discharge of thespent LWR fuel with burnup higher than 18 MWd/kgU can be determined using a simple formula

0 ( )bpS a BU= (2),

where 0pS is the primary 244Cm neutron source at

discharge,BU is the discharge burnup,a and b are fuel bundle type dependent

constants.

4 NEUTRON SOURCE ASSESSMENT

Figure 1. Total neutron source and its major components as a funtion of cooling time(enrichment 2.746 w-%, burnup 37.8 MWd/kgU, and coolant density 0.426 g/cm3).

10

S T U K - Y T O - T R 1 7 0

At Kernforschungszentrum Karlsruhe burnup-de-pendent neutron emission profiles of spent BWRand PWR fuel were measured by a combination ofactive and passive neutron measurement, and themean neutron emission of the fuel bundles weredetermined. The neutron emission values werecorrected for the decay of 244Cm, and the 242Cmcontribution was eliminated. According to Würz bis 3.97 for BWR, and 3.94 for PWR spent fuel; andthe initial enrichment does not affect b [6]. At VTTEnergy Anttila has studied the burnup-depend-ence [5]. His ORIGEN-2 calculations implied thatb varies from 4.5 to 4.8.

The burnup-dependence of the primary 244Cmneutron source at discharge was studied by a serieof SAS2H/ORIGEN-S calculations. Discharge bur-nup varied from 15 to 40 MWd/kgU. The fuelbundle was a common 8x8 BWR fuel bundle with-out burnable absorber rods. A uniform enrichmentof 3.0 w-% was applied, and the moderator densitywas set 0.5. Average power of the bundle was set4.5 MW, and the desired discharge burnup wasachieved by adjusting the time that the bundlewas irradiated in the reactor. The calculated burn-up-dependence of total neutron source, and 242Cmand 244Cm spontaneous fission neutron sourcesare shown in Figure 2. The value of b was deter-mined using least-square-method, and it was

found to be 4.67. The value of a was found to be 5.The least-square fit was good (R2=0.9992), and themodel was capable of predicting the neutronsource within 7% when burnup varied from 15 to40 MWd/kgU. The model tended to overpredictthe neutron source at low and high burnups, andgive too low neutron source strength at intermedi-ate values of burnup. The determination of the bvalue was repeated at moderator density 0.752 g/cm3 (zero void fraction), and it b was found to be5.03. The fact that b does not stay constant whenvoid fraction varies implies that the simple 244Cmneutron source model is not adequate for accurateassessment of the neutron source.

4.3 Enrichment dependence

The enrichment-dependence of the primary 244Cmneutron source at discharge was studied by a serieof SAS2H/ORIGEN-S calculations. The uniformenrichment was varied from 1.6 up to 4.4 w-%.The fuel bundle was a common 8×8 BWR fuelwithout burnable absorber rods. The void fractionwas set 0. The parameter study was performed atthree different discharge burnups: 20, 30 and40 MWd/kgU. The average power of the bundlewas set 4.5 MW, and the desired discharge burnupwas achieved by adjusting the time that the bun-

Figure 2. Burnup-dependence of total neutron source, and 242Cm and 244Cm spontaneous fission neutronsources according to SAS2H/ORIGEN-S calculations.

1,0E+06

1,0E+07

1,0E+08

1,0E+09

10 15 20 25 30 35 40 45

Burnup (MWd/kgU)

Neu

tro

n so

urce

(n/

s/as

sem

bly

)

total

cm-242

cm-244

S T U K - Y T O - T R 1 7 0

11

dle was irradiated in the reactor. The calculatedprimary 244Cm neutron sources at discharge areshown in Table III.

The parameter a in the simple neutron sourcemodel (Equation 2) was calculated by dividing thecomputed primary 244Cm neutron source at dis-charge by (BU)b, where b was 5.03. The enrich-ment dependence of a is shown in Figure 3. Forthe same burnup, lower enrichment gave higherneutron source. The value of the b parameter wasdetermined at 3 w-% enrichment, and the simple244Cm neutron source model was able to give goodresults around and above that enrichment. Thecurves diverge at lower enrichments, which re-veals the inadequacy of the model.

4.4 Void fraction dependence

The void fraction dependence of the primary 244Cmneutron source at discharge was studied by a serieof SAS2H/ORIGEN-S calculations. The coolantvoid fraction inside the assembly varied from 0 to0.8, while the channel void fraction was set 0. Thefuel bundle was a common 8x8 BWR fuel withoutburnable absorber rods. A uniform enrichment of3.0 w-% was applied. The parameter study wascarried out at three different values of dischargeburnup: 20, 30 and 40 MWd/kgU. Average powerof the bundle was set 4.5 MW, and the desireddischarge burnup was achieved by adjusting thetime that the bundle spent in the reactor. The

calculated primary 244Cm neutron sources at dis-charge are shown in Table IV.

The parameter a in the simple 244Cm neutronsource model (Equation 2) was calculated by di-

Figure 3. Enrichment dependence of the parameter a in simple 244Cm neutron source model.

0

0,5

1

1,5

2

2,5

3

3,5

4

4,5

5

1,5 2,0 2,5 3,0 3,5 4,0 4,5

Enrichment (w-%)

a(b

=5.

03, a

=0,

no

bur

nab

le a

bso

rber

rod

s)

20

30

40

Poly. (20)

Poly. (30)

Poly. (40)

Table III. The calculated enrichment dependent244Cm neutron source at discharge (n/s/assembly).

Table IV. The calculated void fraction dependent244Cm neutron source at discharge (n/s/assembly).

tnemhcirnE )Ugk/dWM(punruB

)%-w( 02 03 04

6.1 01·06.1 7 01·38.9 7 01·00.3 8

0.2 01·87.9 6 01·77.6 7 01·03.2 8

4.2 01·32.6 6 01·66.4 7 01·47.1 8

8.2 01·61.4 6 01·72.3 7 01·03.1 8

2.3 01·98.2 6 01·43.2 7 01·27.9 7

6.3 01·90.2 6 01·27.1 7 01·53.7 7

0.4 01·65.1 6 01·03.1 7 01·36.5 7

4.4 01·91.1 6 01·00.1 7 01·93.4 7

)Ugk/dWM(punruB

noitcarfdioV 02 03 04

0.0 01·54.3 6 01·67.2 7 01·21.1 8

2.0 01·25.4 6 01·53.3 7 01·82.1 8

4.0 01·91.6 6 01·22.4 7 01·94.1 8

6.0 01·78.8 6 01·64.5 7 01·77.1 8

8.0 01·23.1 7 01·91.7 7 01·01.2 8

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0

1

2

3

4

0 0,1 0,2 0,3 0,4 0,5 0,6 0,7 0,8

Void fraction

a(e=

3 w

-%, b

=5.

03, n

o b

urna

ble

ab

sorb

er

rod

s)

203040Poly. (20)Poly. (30)Poly. (40)

viding the computed primary 244Cm neutronsource at discharge by (BU)b, where b was 5.03.The void fraction dependence of a is shown inFigure 4. Void fraction had a strong influence onbuildup of the 244Cm in BWR: for the same burnuphigher void fraction gave a higher neutron source.The value of b was determined at zero void frac-tion, where the value of the parameter a was thesame for all three burnups. The curves diverge athigher void fractions, which reveals the inadequa-cy of the simple 244Cm neutron source model.

4.5 Effect of the amount of Gd2O3in burnable absorber rods

The effect of the amount of Gd2O3 in the burnableabsorber rods on the primary 244Cm neutronsource at discharge was studied by a serie ofSAS2H/ORIGEN-S calculations. The fuel bundlewas a common 8x8 BWR fuel bundle with fourburnable absorber rods. The amount of Gd2O3 inburnable absorber rods was varied from 2.0 up to4.0 w-%. The void fraction was set 0. The parame-ter study was performed at three different dis-charge burnups: 20, 30 and 40 MWd/kgU. Averagepower of the bundle was set 4.5 MW, and the de-sired discharge burnup was achieved by adjustingthe time that the bundle was irradiated in thereactor. The calculated primary 244Cm neutronsources at discharge are shown in Table V. Accord-

ing to the results the primary 244Cm neutronsource at discharge is only weakly dependent onthe amount of Gd2O3 in the burnable absorberrods.

4.6 Axial neutron sourcedistribution

The fuel bundle neutron source has an axial dis-tribution because burnup and void fraction varyaxially. The axial distribution of the 244Cm inID6124 BWR fuel assembly was calculated withSAS2H/ORIGEN-S using the node-wise (25 nodesaxially, 0=bottom) burnup and void fraction dataacquired from SIMULATE-3 runs. The ID6124bundle is a 8×8 BWR fuel bundle with four burna-ble absorber rods. Its average enrichment is2.75 w-%, and its average burnup at discharge is

Figure 4. Void fraction dependence of the parameter in a simple 244Cm neutron source model.

Table V. The calculated 244Cm neutron source atdischarge (n/s/assembly).

dG 2O3 sdorABni )Ugk/dWM(punruB

)%-w( 02 03 04

0.2 01·85.3 6 01·08.2 7 01·31.1 8

5.2 01·26.3 6 01·18.2 7 01·31.1 7

0.3 01·46.3 6 01·28.2 7 01·31.1 7

5.3 01·66.3 6 01·38.2 7 01·31.1 7

0.4 01·86.3 6 01·48.2 7 01·31.1 7

S T U K - Y T O - T R 1 7 0

13

37.8 MWd/kgU. The axial variation in gadoliniumcontent was not taken into account in the calcula-tions since the amount of the burnable absorberdoes not significantly influence the magnitude ofthe neutron source. The normalized calculated ax-ial 244Cm distribution and the assumed dischargeburnup of the ID6124 BWR bundle are shown inFigure 5. The figure implies that the higher voidfraction in the upper part of the fuel bundle movesthe maximum axial neutron source towards thetop of the assembly.

4.7 Pinwise neutron sourcedistribution

There is also a radial or more likely pinwise neu-tron source distribution in the fuel bundle. This isdue to the variations in the pin enrichment andthe irradiation conditions (neutron flux) of eachpin. In this work the pinwise distribution of the244Cm concentration was studied by CASMO-4. How-ever, it was realised that the calculation methodwas not valid for studying the pinwise distribu-tion: CASMO-4 is a fuel assembly burnup codethat assumes reflecting boundary conditions at thebundle boundaries. The surrounding fuel bundlesor control rods affect the buildup of neutron source.In order to soundly assess the pinwise neutronsource distribution one should use a core simula-tor coupled with a pin power reconstruction model.

4.8 Assessment of the neutronsource of the Olkiluoto BWRbundles

The neutron source of nine Olkiluoto BWR fuelbundles were calculated with SAS2H/ORIGEN-Susing as exact models and input data as possible.The input data included data about the fuel bun-dle type, average burnup and average void frac-tion from SIMULATE-3 runs and cooling time.The neutron emission of the nine fuel bundleswere measured in November 1998 under TaskJNT 1077 in the KPA Store in Olkiluoto with IAEABWR Fork detector coupled with a compact under-water gamma spectrometer. The measured neu-tron emissions are plotted against calculationallyassessed neutron sources in Figure 6. There is aclear correlation between the measured neutronemission and the calculated neutron source. How-ever, to be precise the multiplication of the neu-trons in the fuel matrix should also be taken intoaccount. The isotopic contents of the spent fuelaffect the neutron emission and the effectivesource in a subcritical source is

1p

eff

SS

k=

-

(3),

where Sp is the primary 244Cm neutron source,k is the multiplication factor of the

system.

0 %

20 %

40 %

60 %

80 %

100 %

120 %

140 %

0 5 10 15 20 25

Node

No

rmal

ized

Axi

al C

m-2

44

Co

ncen

trat

ion

and

Bur

nup

Cm-244 Concentration Discharge Burnup

Poly. (Discharge Burnup) Poly. (Cm-244 Concentration)

Figure 5. The calculated axial distribution of the Cm-244 concentration and assumed discharge burnup.

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S T U K - Y T O - T R 1 7 0

Figure 6. Measured neutron emission vs. calculationally assessed neutron source.

Figure 7. Amount of 137Cs as a function of burnup according to SAS2H/ORIGEN-S.

0

50

100

150

200

250

300

10 15 20 25 30 35 40 45

Burnup (MWd/kgU)

Cs-

137

(gra

ms)

0

1000

2000

3000

4000

0 2 4 6 8 10 12 14

Calculated Neutron Source (1E7 n/s/assembly)

Mea

sure

d N

eutr

on

Em

issi

on

ID6124

ID7516

ID12323

ID426

ID104

ID7472

ID9367

ID12350

ID12338

S T U K - Y T O - T R 1 7 0

15

Spent LWR fuel contains a variety of gamma emit-ting nuclides (Appendix D). However, only 137Csand 134Cs gamma peaks stood out from the gammaspectrum measured during the underwater gam-ma spectrometry measurement campaign [2]. Thefission product contents of the spent fuel is closelycorrelated to its burnup. 137Cs is mainly produceddirectly from fission but 134Cs is mainly producedfrom neutron reaction with 133Cs.

The SAS2H/ORIGEN-S runs performed tostudy the neutron source of spent BWR fuel alsoproduced information about the 137Cs and 134Cscontents of the fuel. As expected, the only signifi-cant relations between the amount of the cesiumisotopes and the parameters studied were those ofburnup and cooling time decay (no relation be-

5 ASSESSMENT OF THE 137Cs AND 134CsGAMMA SOURCES

tween the amount of the cesium isotopes andenrichment, void fraction or the amount of burna-ble absorber). The calculated 137Cs and 134Cs con-tents of the spent BWR fuel are shown in Figures7 and 8. The results imply that 137Cs is a very goodindicator of the fuel burnup; and that 134Cs is notas good because its production increases at higherburnups. Anyhow, the ratio between the amountof 134Cs and 137Cs can be used to estimate thecooling time Tc according to formula

03.198 ln( )cc

rT

r= (4),

where r0 is the ratio between the amount of134Cs and 137Cs at discharge,

rc is the ratio after Tc years of cooling.

Figure 8. Amount of 134Cs as a function of burnup according to SAS2H/ORIGEN-S.

0

5

10

15

20

25

30

10 15 20 25 30 35 40 45

Burnup (MWd/kgU)

Cs-

134

(gra

ms)

16

S T U K - Y T O - T R 1 7 0

SAS2H/ORIGEN-S was succesfully applied to as-sess the neutron and the gamma sources of thespent BWR and VVER-440 fuel. CASMO-4 wasapplied as a reference code to estimate the neu-tron source of the BWR fuel. The parameter stud-ies performed gave valuable information aboutwhat has to be taken into account in order to accu-rately assess the neutron and gamma sources ofthe spent LWR fuel.

The neutron emission of the spent LWR fuel isdominantly from spontaneous fission of 244Cm atcooling times longer than two years. Because ofthe long half-life of 244Cm, neutron emission re-mains a characteristic feature of the spent LWRfuel even after relatively long cooling times. Theparameters that affect the 244Cm concentration atdischarge are fuel bundle type, burnup, and incase of BWR also void fraction. The power historyhas only a slight influence on buildup of 244Cm.The axial distribution of the neutron source can bedetermined if all the required input data is availa-ble. The accurate determination of the pinwiseneutron source distribution is a demanding task.

Spent LWR fuel contains a variety of gammaemitting nuclides. However, only 137Cs and 134Csgamma peaks stand out from the gamma spec-

trum. According to the SAS2H/ORIGEN-S results137Cs is a better indicator of the burnup than 134Cs.Anyhow, the ratio between the amount of 134Csand 137Cs can be used to estimate the cooling time.

Modern BWR fuel assemblies are highly heter-ogeneous: there are pins with different enrich-ments, burnable absorbers are used, and alsoaxial variations (for example part length fuel rods,axial variation in the amount of burnable absorb-er, natural uranium end parts). Moreover, controlrods affect the way the fuel is burnt in the reactor.Thus, assessment of the radiation source distribu-tion of BWR fuel bundle is a relatively complicat-ed task and some uncertainty is to be accepted.

When neutron or gamma radiation is beingdetected one should take into account the fact thatin addition to the primary source particles thereare secondary particles induced by the primaryones (for example, the fuel matrix of the spent fuelis acting as a neutron multiplier). Thus, for soundinterpretation of the detector signals the produc-tion of the secondary particles should also beconsidered. This requires knowledge of the spatialsource distribution, and rigorous modelling of boththe spent fuel matrix and the detection geometry.

6 CONCLUSIONS

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17

[1] SCALE-4.3, Modular System for PerformingStandardized Computer Analyses for Licenc-ing Evaluation. Oak Ridge National Laborato-ry, 1997. RSIC Computer Code CollectionCCC-0545/12.

[2] Tiitta A, Hautamäki J, Turunen A., Arlt R,Arenas Carrasco J, Esmailpour-Kazerouni K,Schwalbach P. Underwater Gamma Spectrom-etry in Conjunction with the Fork Detector,Technical report on the measurements con-ducted with an upgraded fork detector at theOlkiluoto KPA store 1.–3.9.1999. VTT Chemi-cal Technology research report, November1999.

[3] Tarvainen M, Bäcklin A, Håkansson A. Cali-bration of the TVO spent BWR reference fuelassembly, Final report on the joint TaskJNT61 of the Finnish and Swedish SupportProgrammes to IAEA Safeguards. FinnishCentre for Radiation and Nuclear Safety,STUK-YTO-TR 37. 1992.

REFERENCES

[4] Mandoki R, Bruggeman M, Baeten P, CarchonR. A Feasibility Study of Passive NDA Tech-niques for the Verification of the HLW GlassCanister by Monte Carlo Simulations. SCK/CEN Research Report BLG-785, August 1998.

[5] Anttila M. Gamma and neutron dose rates onthe outer surface of the nuclear waste disposalcanisters. Posiva report 96-10, 1996.

[6] Würz H. A simple nondestructive measure-ment system for spent-fuel management. Nu-clear Technology, Vol. 95, 1991, pp. 193–206.

18

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APPENDIX A NUCLIDE INVENTORY OF SPENT BWR FUELbu= 37816, bwr, 5 cyc, e=2.746, vf=0.426, 8x8-e2-275-1-200-10actinidesdecay, following reactor irradiation identified by: power=4.31mw, burnup=6299.mwd, flux=3.43E+13n/cm**2-sec nuclide concentrations, grams basis =single reactor assembly

initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d he 4 1.45E-01 2.32E-01 2.66E-01 2.99E-01 3.31E-01 3.62E-01 3.94E-01 pb208 1.23E-07 6.79E-07 1.97E-06 3.93E-06 6.35E-06 9.07E-06 1.20E-05 th228 3.36E-07 1.24E-06 2.23E-06 2.99E-06 3.50E-06 3.80E-06 3.96E-06 th230 1.65E-06 6.42E-06 1.54E-05 2.86E-05 4.59E-05 6.71E-05 9.24E-05 th232 5.50E-05 9.51E-05 1.35E-04 1.75E-04 2.15E-04 2.56E-04 2.96E-04 th234 2.28E-06 2.28E-06 2.28E-06 2.28E-06 2.28E-06 2.28E-06 2.28E-06 pa231 3.98E-06 5.64E-06 7.29E-06 8.94E-06 1.06E-05 1.22E-05 1.39E-05 pa233 2.73E-06 2.74E-06 2.75E-06 2.76E-06 2.77E-06 2.78E-06 2.79E-06 u232 3.97E-05 9.38E-05 1.24E-04 1.40E-04 1.48E-04 1.51E-04 1.52E-04 u233 7.17E-05 1.32E-04 1.92E-04 2.53E-04 3.14E-04 3.75E-04 4.36E-04 u234 4.35E-01 1.11E+00 1.79E+00 2.45E+00 3.11E+00 3.75E+00 4.38E+00 u235 7.60E+02 7.60E+02 7.60E+02 7.60E+02 7.60E+02 7.60E+02 7.60E+02 u236 6.15E+02 6.15E+02 6.15E+02 6.15E+02 6.15E+02 6.15E+02 6.16E+02 u237 1.19E+00 6.49E-06 5.82E-06 5.22E-06 4.69E-06 4.21E-06 3.78E-06 u238 1.57E+05 1.57E+05 1.57E+05 1.57E+05 1.57E+05 1.57E+05 1.57E+05 np236 1.38E-04 1.38E-04 1.38E-04 1.38E-04 1.38E-04 1.38E-04 1.38E-04 np237 7.95E+01 8.08E+01 8.10E+01 8.12E+01 8.15E+01 8.19E+01 8.23E+01 np239 1.15E+01 2.74E-05 2.73E-05 2.73E-05 2.73E-05 2.73E-05 2.73E-05 pu236 1.36E-04 7.99E-05 4.68E-05 2.74E-05 1.60E-05 9.38E-06 5.49E-06 pu238 3.64E+01 3.92E+01 3.86E+01 3.79E+01 3.72E+01 3.66E+01 3.60E+01 pu239 8.84E+02 8.95E+02 8.95E+02 8.95E+02 8.95E+02 8.95E+02 8.95E+02 pu240 4.56E+02 4.56E+02 4.57E+02 4.58E+02 4.59E+02 4.59E+02 4.60E+02 pu241 2.38E+02 2.14E+02 1.92E+02 1.72E+02 1.55E+02 1.39E+02 1.25E+02 pu242 1.27E+02 1.27E+02 1.27E+02 1.27E+02 1.27E+02 1.27E+02 1.27E+02 am241 1.05E+01 3.49E+01 5.66E+01 7.61E+01 9.35E+01 1.09E+02 1.23E+02 am242m 2.13E-01 2.11E-01 2.08E-01 2.06E-01 2.04E-01 2.02E-01 1.99E-01 am242 2.14E-02 2.72E-06 2.69E-06 2.66E-06 2.63E-06 2.60E-06 2.57E-06 am243 3.18E+01 3.18E+01 3.18E+01 3.18E+01 3.18E+01 3.18E+01 3.18E+01 cm242 3.37E+00 1.05E-01 3.75E-03 6.37E-04 5.35E-04 5.26E-04 5.20E-04 cm243 1.09E-01 1.03E-01 9.75E-02 9.23E-02 8.74E-02 8.28E-02 7.84E-02 cm244 1.23E+01 1.13E+01 1.03E+01 9.49E+00 8.71E+00 7.99E+00 7.33E+00 cm245 4.57E-01 4.57E-01 4.57E-01 4.57E-01 4.56E-01 4.56E-01 4.56E-01 cm246 8.85E-02 8.85E-02 8.85E-02 8.84E-02 8.84E-02 8.84E-02 8.83E-02 cm247 1.85E-03 1.85E-03 1.85E-03 1.85E-03 1.85E-03 1.85E-03 1.85E-03 cm248 2.18E-04 2.18E-04 2.18E-04 2.18E-04 2.18E-04 2.18E-04 2.18E-04 cf249 6.02E-07 3.57E-06 4.06E-06 4.13E-06 4.13E-06 4.11E-06 4.09E-06total 1.61E+05 1.61E+05 1.61E+05 1.61E+05 1.61E+05 1.61E+05 1.61E+05

bu=37816, bwr, 5 cyc, e=2.746, vf=0.426, 8x8-e2-275-1-200-10fission productsdecay, following reactor irradiation identified by: power=4.31mw, burnup=6299.mwd, flux=3.43E+13n/cm**2-sec nuclide concentrations, grams basis = single reactor assembly h 3 9.17E-03 8.09E-03 7.13E-03 6.29E-03 5.54E-03 4.89E-03 4.31E-03 li 6 2.85E-05 2.85E-05 2.85E-05 2.85E-05 2.85E-05 2.85E-05 2.85E-05 li 7 1.92E-06 1.92E-06 1.92E-06 1.92E-06 1.92E-06 1.92E-06 1.92E-06 be 9 3.71E-06 3.71E-06 3.71E-06 3.71E-06 3.71E-06 3.71E-06 3.71E-06 be 10 2.47E-05 2.47E-05 2.47E-05 2.47E-05 2.47E-05 2.47E-05 2.47E-05 c 14 5.00E-06 5.00E-06 4.99E-06 4.99E-06 4.99E-06 4.99E-06 4.99E-06 zn 70 1.80E-06 1.80E-06 1.80E-06 1.80E-06 1.80E-06 1.80E-06 1.80E-06 ga 71 1.71E-05 1.71E-05 1.71E-05 1.71E-05 1.71E-05 1.71E-05 1.71E-05 ge 72 1.01E-03 1.02E-03 1.02E-03 1.02E-03 1.02E-03 1.02E-03 1.02E-03 ge 73 2.84E-03 2.84E-03 2.84E-03 2.84E-03 2.84E-03 2.84E-03 2.84E-03 ge 74 2.49E-03 2.49E-03 2.49E-03 2.49E-03 2.49E-03 2.49E-03 2.49E-03 as 75 2.11E-02 2.11E-02 2.11E-02 2.11E-02 2.11E-02 2.11E-02 2.11E-02 ge 76 6.28E-02 6.28E-02 6.28E-02 6.28E-02 6.28E-02 6.28E-02 6.28E-02 se 76 8.14E-04 8.16E-04 8.16E-04 8.16E-04 8.16E-04 8.16E-04 8.16E-04 se 77 1.41E-01 1.41E-01 1.41E-01 1.41E-01 1.41E-01 1.41E-01 1.41E-01 se 78 4.93E-01 4.93E-01 4.93E-01 4.93E-01 4.93E-01 4.93E-01 4.93E-01 se 79 9.11E-01 9.11E-01 9.11E-01 9.11E-01 9.11E-01 9.11E-01 9.11E-01 br 79 4.79E-06 9.08E-06 1.34E-05 1.77E-05 2.20E-05 2.62E-05 3.05E-05 se 80 2.48E+00 2.48E+00 2.48E+00 2.48E+00 2.48E+00 2.48E+00 2.48E+00 kr 80 1.24E-05 1.24E-05 1.24E-05 1.24E-05 1.24E-05 1.24E-05 1.24E-05 br 81 3.66E+00 3.66E+00 3.66E+00 3.66E+00 3.66E+00 3.66E+00 3.66E+00 kr 81 1.05E-06 1.05E-06 1.05E-06 1.05E-06 1.05E-06 1.05E-06 1.05E-06 se 82 5.87E+00 5.87E+00 5.87E+00 5.87E+00 5.87E+00 5.87E+00 5.87E+00 kr 82 1.42E-01 1.42E-01 1.42E-01 1.42E-01 1.42E-01 1.42E-01 1.42E-01 kr 83 6.83E+00 6.83E+00 6.83E+00 6.83E+00 6.83E+00 6.83E+00 6.83E+00 kr 84 2.11E+01 2.11E+01 2.11E+01 2.11E+01 2.11E+01 2.11E+01 2.11E+01

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NUCLIDE INVENTORY OF SPENT BWR FUEL APPENDIX A kr 85 3.76E+00 3.25E+00 2.81E+00 2.43E+00 2.11E+00 1.82E+00 1.58E+00 rb 85 1.67E+01 1.72E+01 1.76E+01 1.80E+01 1.83E+01 1.86E+01 1.88E+01 kr 86 3.25E+01 3.25E+01 3.25E+01 3.25E+01 3.25E+01 3.25E+01 3.25E+01 sr 86 8.13E-02 8.42E-02 8.42E-02 8.42E-02 8.42E-02 8.42E-02 8.42E-02 rb 87 4.23E+01 4.23E+01 4.23E+01 4.23E+01 4.23E+01 4.23E+01 4.23E+01 sr 87 4.80E-04 4.80E-04 4.80E-04 4.80E-04 4.80E-04 4.80E-04 4.80E-04 sr 88 6.05E+01 6.05E+01 6.05E+01 6.05E+01 6.05E+01 6.05E+01 6.05E+01 y 89 7.77E+01 8.04E+01 8.04E+01 8.04E+01 8.04E+01 8.04E+01 8.04E+01 sr 90 9.21E+01 8.72E+01 8.25E+01 7.81E+01 7.39E+01 6.99E+01 6.62E+01 y 90 2.48E-02 2.27E-02 2.14E-02 2.03E-02 1.92E-02 1.82E-02 1.72E-02 zr 90 6.41E+00 1.14E+01 1.60E+01 2.05E+01 2.47E+01 2.86E+01 3.24E+01 zr 91 1.01E+02 1.06E+02 1.06E+02 1.06E+02 1.06E+02 1.06E+02 1.06E+02 zr 92 1.14E+02 1.14E+02 1.14E+02 1.14E+02 1.14E+02 1.14E+02 1.14E+02 zr 93 8.51E+01 8.52E+01 8.52E+01 8.52E+01 8.52E+01 8.52E+01 8.52E+01 nb 93 9.17E-06 2.17E-05 4.11E-05 6.66E-05 9.77E-05 1.34E-04 1.75E-04 nb 93m 9.31E-05 1.67E-04 2.34E-04 2.95E-04 3.50E-04 4.01E-04 4.46E-04 zr 94 1.41E+02 1.41E+02 1.41E+02 1.41E+02 1.41E+02 1.41E+02 1.41E+02 nb 94 9.77E-05 9.77E-05 9.77E-05 9.77E-05 9.77E-05 9.77E-05 9.77E-05 mo 95 1.25E+02 1.37E+02 1.37E+02 1.37E+02 1.37E+02 1.37E+02 1.37E+02 zr 96 1.49E+02 1.49E+02 1.49E+02 1.49E+02 1.49E+02 1.49E+02 1.49E+02 mo 96 9.51E+00 9.51E+00 9.51E+00 9.51E+00 9.51E+00 9.51E+00 9.51E+00 mo 97 1.42E+02 1.42E+02 1.42E+02 1.42E+02 1.42E+02 1.42E+02 1.42E+02 mo 98 1.56E+02 1.56E+02 1.56E+02 1.56E+02 1.56E+02 1.56E+02 1.56E+02 tc 98 1.35E-03 1.35E-03 1.35E-03 1.35E-03 1.35E-03 1.35E-03 1.35E-03 tc 99 1.45E+02 1.45E+02 1.45E+02 1.45E+02 1.45E+02 1.45E+02 1.45E+02 ru 99 6.28E-03 7.36E-03 8.43E-03 9.50E-03 1.06E-02 1.16E-02 1.27E-02 mo100 1.78E+02 1.78E+02 1.78E+02 1.78E+02 1.78E+02 1.78E+02 1.78E+02 ru100 2.48E+01 2.48E+01 2.48E+01 2.48E+01 2.48E+01 2.48E+01 2.48E+01 ru101 1.46E+02 1.46E+02 1.46E+02 1.46E+02 1.46E+02 1.46E+02 1.46E+02 ru102 1.53E+02 1.53E+02 1.53E+02 1.53E+02 1.53E+02 1.53E+02 1.53E+02 rh102 2.16E-04 1.26E-04 7.40E-05 4.33E-05 2.54E-05 1.48E-05 8.69E-06 rh103 7.98E+01 8.57E+01 8.57E+01 8.57E+01 8.57E+01 8.57E+01 8.57E+01 ru104 1.13E+02 1.13E+02 1.13E+02 1.13E+02 1.13E+02 1.13E+02 1.13E+02 pd104 5.44E+01 5.44E+01 5.44E+01 5.44E+01 5.44E+01 5.44E+01 5.44E+01 pd105 8.20E+01 8.22E+01 8.22E+01 8.22E+01 8.22E+01 8.22E+01 8.22E+01 ru106 2.68E+01 5.82E+00 1.26E+00 2.75E-01 5.97E-02 1.30E-02 2.82E-03 pd106 5.41E+01 7.51E+01 7.96E+01 8.06E+01 8.08E+01 8.09E+01 8.09E+01 pd107 5.02E+01 5.02E+01 5.02E+01 5.02E+01 5.02E+01 5.02E+01 5.02E+01 ag107 9.41E-06 2.14E-05 3.34E-05 4.54E-05 5.74E-05 6.94E-05 8.14E-05 pd108 3.31E+01 3.31E+01 3.31E+01 3.31E+01 3.31E+01 3.31E+01 3.31E+01 ag108m 4.74E-05 4.68E-05 4.62E-05 4.57E-05 4.51E-05 4.46E-05 4.40E-05 cd108 4.88E-05 4.89E-05 4.89E-05 4.89E-05 4.90E-05 4.90E-05 4.91E-05 ag109 1.89E+01 1.89E+01 1.89E+01 1.89E+01 1.89E+01 1.89E+01 1.89E+01 pd110 9.91E+00 9.91E+00 9.91E+00 9.91E+00 9.91E+00 9.91E+00 9.91E+00 ag110m 1.62E-01 1.67E-02 1.72E-03 1.78E-04 1.83E-05 1.89E-06 1.95E-07 cd110 1.08E+01 1.09E+01 1.10E+01 1.10E+01 1.10E+01 1.10E+01 1.10E+01 cd111 5.12E+00 5.17E+00 5.17E+00 5.17E+00 5.17E+00 5.17E+00 5.17E+00 cd112 2.62E+00 2.62E+00 2.62E+00 2.62E+00 2.62E+00 2.62E+00 2.62E+00 cd113 1.45E-02 1.49E-02 1.49E-02 1.49E-02 1.49E-02 1.49E-02 1.49E-02 cd113m 2.85E-02 2.55E-02 2.29E-02 2.05E-02 1.83E-02 1.64E-02 1.47E-02 in113 2.39E-03 5.36E-03 8.01E-03 1.04E-02 1.25E-02 1.44E-02 1.61E-02 cd114 2.57E+00 2.57E+00 2.57E+00 2.57E+00 2.57E+00 2.57E+00 2.57E+00 sn114 1.66E-04 1.85E-04 1.85E-04 1.85E-04 1.85E-04 1.85E-04 1.85E-04 in115 2.57E-01 2.61E-01 2.61E-01 2.61E-01 2.61E-01 2.61E-01 2.61E-01 sn115 3.60E-02 3.61E-02 3.61E-02 3.61E-02 3.61E-02 3.61E-02 3.61E-02 cd116 1.01E+00 1.01E+00 1.01E+00 1.01E+00 1.01E+00 1.01E+00 1.01E+00 sn116 5.19E-01 5.19E-01 5.19E-01 5.19E-01 5.19E-01 5.19E-01 5.19E-01 sn117 9.77E-01 9.78E-01 9.78E-01 9.78E-01 9.78E-01 9.78E-01 9.78E-01 sn118 7.71E-01 7.71E-01 7.71E-01 7.71E-01 7.71E-01 7.71E-01 7.71E-01 sn119 8.07E-01 8.10E-01 8.11E-01 8.11E-01 8.11E-01 8.11E-01 8.11E-01 sn120 7.88E-01 7.88E-01 7.88E-01 7.88E-01 7.88E-01 7.88E-01 7.88E-01 sn121m 9.07E-03 8.82E-03 8.57E-03 8.33E-03 8.10E-03 7.87E-03 7.66E-03 sb121 7.70E-01 7.71E-01 7.71E-01 7.71E-01 7.72E-01 7.72E-01 7.72E-01 sn122 1.02E+00 1.02E+00 1.02E+00 1.02E+00 1.02E+00 1.02E+00 1.02E+00 te122 6.37E-02 6.41E-02 6.41E-02 6.41E-02 6.41E-02 6.41E-02 6.41E-02 sb123 9.04E-01 9.13E-01 9.13E-01 9.13E-01 9.13E-01 9.13E-01 9.13E-01 te123 5.90E-04 7.60E-04 7.61E-04 7.61E-04 7.61E-04 7.61E-04 7.61E-04 sn124 1.71E+00 1.71E+00 1.71E+00 1.71E+00 1.71E+00 1.71E+00 1.71E+00 te124 5.28E-02 5.97E-02 5.97E-02 5.97E-02 5.97E-02 5.97E-02 5.97E-02 sb125 1.31E+00 7.44E-01 4.21E-01 2.38E-01 1.35E-01 7.64E-02 4.33E-02 te125 7.87E-01 1.36E+00 1.69E+00 1.88E+00 1.98E+00 2.04E+00 2.07E+00 te125m 1.70E-02 1.06E-02 5.98E-03 3.39E-03 1.92E-03 1.09E-03 6.15E-04 sn126 4.06E+00 4.05E+00 4.05E+00 4.05E+00 4.05E+00 4.05E+00 4.05E+00 te126 8.15E-02 8.22E-02 8.23E-02 8.23E-02 8.24E-02 8.25E-02 8.25E-02 i127 8.50E+00 8.73E+00 8.73E+00 8.73E+00 8.73E+00 8.73E+00 8.73E+00 te128 1.79E+01 1.79E+01 1.79E+01 1.79E+01 1.79E+01 1.79E+01 1.79E+01

20

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APPENDIX A NUCLIDE INVENTORY OF SPENT BWR FUEL xe128 6.55E-01 6.55E-01 6.55E-01 6.55E-01 6.55E-01 6.55E-01 6.55E-01 i129 3.51E+01 3.54E+01 3.54E+01 3.54E+01 3.54E+01 3.54E+01 3.54E+01 xe129 4.93E-03 5.01E-03 5.02E-03 5.02E-03 5.02E-03 5.03E-03 5.03E-03 te130 6.98E+01 6.98E+01 6.98E+01 6.98E+01 6.98E+01 6.98E+01 6.98E+01 xe130 1.84E+00 1.84E+00 1.84E+00 1.84E+00 1.84E+00 1.84E+00 1.84E+00 xe131 7.18E+01 7.28E+01 7.28E+01 7.28E+01 7.28E+01 7.28E+01 7.28E+01 xe132 2.18E+02 2.19E+02 2.19E+02 2.19E+02 2.19E+02 2.19E+02 2.19E+02 ba132 4.45E-05 4.51E-05 4.51E-05 4.51E-05 4.51E-05 4.51E-05 4.51E-05 cs133 2.09E+02 2.11E+02 2.11E+02 2.11E+02 2.11E+02 2.11E+02 2.11E+02 xe134 2.86E+02 2.86E+02 2.86E+02 2.86E+02 2.86E+02 2.86E+02 2.86E+02 cs134 2.28E+01 1.07E+01 5.05E+00 2.38E+00 1.12E+00 5.28E-01 2.48E-01 ba134 1.32E+01 2.53E+01 3.09E+01 3.36E+01 3.49E+01 3.55E+01 3.57E+01 cs135 7.53E+01 7.53E+01 7.53E+01 7.53E+01 7.53E+01 7.53E+01 7.53E+01 ba135 1.26E-01 1.27E-01 1.27E-01 1.27E-01 1.27E-01 1.27E-01 1.27E-01 xe136 4.30E+02 4.30E+02 4.30E+02 4.30E+02 4.30E+02 4.30E+02 4.30E+02 ba136 4.30E+00 4.40E+00 4.40E+00 4.40E+00 4.40E+00 4.40E+00 4.40E+00 cs137 2.31E+02 2.19E+02 2.08E+02 1.98E+02 1.88E+02 1.78E+02 1.69E+02 ba137 1.30E+01 2.47E+01 3.58E+01 4.63E+01 5.62E+01 6.57E+01 7.47E+01 ba137m 3.55E-05 3.35E-05 3.18E-05 3.02E-05 2.87E-05 2.72E-05 2.58E-05 ba138 2.41E+02 2.41E+02 2.41E+02 2.41E+02 2.41E+02 2.41E+02 2.41E+02 la138 1.68E-03 1.68E-03 1.68E-03 1.68E-03 1.68E-03 1.68E-03 1.68E-03 la139 2.26E+02 2.26E+02 2.26E+02 2.26E+02 2.26E+02 2.26E+02 2.26E+02 ce140 2.42E+02 2.45E+02 2.45E+02 2.45E+02 2.45E+02 2.45E+02 2.45E+02 pr141 2.02E+02 2.08E+02 2.08E+02 2.08E+02 2.08E+02 2.08E+02 2.08E+02 ce142 2.11E+02 2.11E+02 2.11E+02 2.11E+02 2.11E+02 2.11E+02 2.11E+02 nd142 4.82E+00 4.83E+00 4.83E+00 4.83E+00 4.83E+00 4.83E+00 4.83E+00 nd143 1.33E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 ce144 4.19E+01 5.72E+00 7.81E-01 1.07E-01 1.46E-02 1.99E-03 2.72E-04 nd144 2.13E+02 2.49E+02 2.54E+02 2.55E+02 2.55E+02 2.55E+02 2.55E+02 nd145 1.21E+02 1.21E+02 1.21E+02 1.21E+02 1.21E+02 1.21E+02 1.21E+02 pm145 4.00E-06 3.93E-06 3.65E-06 3.36E-06 3.08E-06 2.82E-06 2.58E-06 nd146 1.33E+02 1.33E+02 1.33E+02 1.33E+02 1.33E+02 1.33E+02 1.33E+02 pm146 1.24E-03 9.40E-04 7.09E-04 5.36E-04 4.05E-04 3.06E-04 2.31E-04 sm146 1.73E-03 1.83E-03 1.91E-03 1.97E-03 2.01E-03 2.05E-03 2.07E-03 pm147 2.49E+01 1.43E+01 7.90E+00 4.37E+00 2.42E+00 1.34E+00 7.40E-01 sm147 1.70E+01 2.85E+01 3.49E+01 3.84E+01 4.04E+01 4.15E+01 4.20E+01 nd148 6.92E+01 6.92E+01 6.92E+01 6.92E+01 6.92E+01 6.92E+01 6.92E+01 sm148 2.72E+01 2.75E+01 2.75E+01 2.75E+01 2.75E+01 2.75E+01 2.75E+01 sm149 3.08E-01 4.80E-01 4.80E-01 4.80E-01 4.80E-01 4.80E-01 4.80E-01 nd150 3.49E+01 3.49E+01 3.49E+01 3.49E+01 3.49E+01 3.49E+01 3.49E+01 sm150 6.11E+01 6.11E+01 6.11E+01 6.11E+01 6.11E+01 6.11E+01 6.11E+01 sm151 2.66E+00 2.64E+00 2.60E+00 2.55E+00 2.51E+00 2.47E+00 2.43E+00 eu151 3.31E-03 4.93E-02 9.46E-02 1.39E-01 1.83E-01 2.26E-01 2.68E-01 sm152 2.41E+01 2.41E+01 2.41E+01 2.41E+01 2.41E+01 2.41E+01 2.41E+01 eu152 8.08E-03 7.19E-03 6.40E-03 5.70E-03 5.07E-03 4.51E-03 4.02E-03 gd152 1.47E-02 1.50E-02 1.52E-02 1.54E-02 1.56E-02 1.57E-02 1.59E-02 eu153 2.46E+01 2.48E+01 2.48E+01 2.48E+01 2.48E+01 2.48E+01 2.48E+01 sm154 7.71E+00 7.71E+00 7.71E+00 7.71E+00 7.71E+00 7.71E+00 7.71E+00 eu154 4.94E+00 4.12E+00 3.44E+00 2.87E+00 2.40E+00 2.00E+00 1.67E+00 gd154 6.12E-01 1.43E+00 2.11E+00 2.68E+00 3.15E+00 3.55E+00 3.88E+00 eu155 1.15E+00 8.24E-01 5.91E-01 4.24E-01 3.05E-01 2.19E-01 1.57E-01 gd155 1.10E-02 3.35E-01 5.68E-01 7.35E-01 8.55E-01 9.41E-01 1.00E+00 gd156 1.98E+01 2.05E+01 2.05E+01 2.05E+01 2.05E+01 2.05E+01 2.05E+01 gd157 1.75E-02 2.03E-02 2.03E-02 2.03E-02 2.03E-02 2.03E-02 2.03E-02 gd158 4.16E+00 4.16E+00 4.16E+00 4.16E+00 4.16E+00 4.16E+00 4.16E+00 tb159 5.14E-01 5.15E-01 5.15E-01 5.15E-01 5.15E-01 5.15E-01 5.15E-01 gd160 2.33E-01 2.33E-01 2.33E-01 2.33E-01 2.33E-01 2.33E-01 2.33E-01 dy160 5.29E-02 6.82E-02 6.83E-02 6.83E-02 6.83E-02 6.83E-02 6.83E-02 dy161 7.49E-02 7.63E-02 7.63E-02 7.63E-02 7.63E-02 7.63E-02 7.63E-02 dy162 6.25E-02 6.25E-02 6.25E-02 6.25E-02 6.25E-02 6.25E-02 6.25E-02 dy163 5.52E-02 5.52E-02 5.52E-02 5.52E-02 5.52E-02 5.52E-02 5.52E-02 dy164 1.19E-02 1.19E-02 1.19E-02 1.19E-02 1.19E-02 1.19E-02 1.19E-02 ho165 2.08E-02 2.08E-02 2.08E-02 2.08E-02 2.08E-02 2.08E-02 2.08E-02 ho166m 8.19E-05 8.18E-05 8.17E-05 8.16E-05 8.15E-05 8.14E-05 8.13E-05 er166 5.18E-03 5.20E-03 5.20E-03 5.20E-03 5.20E-03 5.20E-03 5.20E-03 er167 1.06E-04 1.06E-04 1.06E-04 1.06E-04 1.06E-04 1.06E-04 1.06E-04 er168 1.70E-04 1.70E-04 1.70E-04 1.70E-04 1.70E-04 1.70E-04 1.70E-04 yb171 4.49E-07 7.71E-07 9.14E-07 9.78E-07 1.01E-06 1.02E-06 1.03E-06total 6.45E+03 6.45E+03 6.45E+03 6.45E+03 6.45E+03 6.45E+03 6.45E+03

S T U K - Y T O - T R 1 7 0

21

NUCLIDE INVENTORY OF SPENT VVER-440 FUEL APPENDIX Bvver440 fuel bu= 36 mwd/kguactinidesdecay, following reactor irradiation identified by: power= 4.96mw, burnup= 4464.mwd, flux=4.08E+13n/cm**2-sec nuclide concentrations, grams basis = single reactor assembly

initial 60.8 d 121.7 d 182.5 d 243.3 d 304.2 d 365.0 d

he 4 3.77E-02 4.55E-02 5.18E-02 5.69E-02 6.11E-02 6.47E-02 6.76E-02 th230 6.58E-07 7.69E-07 8.93E-07 1.03E-06 1.18E-06 1.34E-06 1.52E-06 th232 2.74E-05 3.01E-05 3.28E-05 3.55E-05 3.82E-05 4.09E-05 4.36E-05 th234 1.69E-06 1.69E-06 1.69E-06 1.69E-06 1.69E-06 1.69E-06 1.69E-06 pa231 3.30E-06 3.50E-06 3.69E-06 3.88E-06 4.08E-06 4.27E-06 4.47E-06 pa233 1.98E-06 2.07E-06 2.09E-06 2.09E-06 2.09E-06 2.09E-06 2.09E-06 u232 1.76E-05 2.13E-05 2.49E-05 2.83E-05 3.15E-05 3.46E-05 3.76E-05 u233 6.82E-05 7.14E-05 7.47E-05 7.79E-05 8.12E-05 8.45E-05 8.77E-05 u234 2.27E-01 2.54E-01 2.82E-01 3.10E-01 3.38E-01 3.66E-01 3.95E-01 u235 1.17E+03 1.17E+03 1.17E+03 1.17E+03 1.17E+03 1.17E+03 1.17E+03 u236 5.58E+02 5.58E+02 5.58E+02 5.58E+02 5.58E+02 5.58E+02 5.58E+02 u237 1.47E+00 2.85E-03 1.08E-05 5.21E-06 5.16E-06 5.12E-06 5.08E-06 u238 1.17E+05 1.17E+05 1.17E+05 1.17E+05 1.17E+05 1.17E+05 1.17E+05 np236 8.70E-05 8.70E-05 8.70E-05 8.70E-05 8.70E-05 8.70E-05 8.70E-05 np237 6.02E+01 6.17E+01 6.17E+01 6.17E+01 6.17E+01 6.17E+01 6.17E+01 np239 1.23E+01 1.27E-05 1.25E-05 1.25E-05 1.25E-05 1.25E-05 1.25E-05 pu236 9.69E-05 9.35E-05 8.99E-05 8.64E-05 8.30E-05 7.98E-05 7.66E-05 pu238 2.06E+01 2.12E+01 2.15E+01 2.17E+01 2.19E+01 2.20E+01 2.21E+01 pu239 7.19E+02 7.31E+02 7.31E+02 7.31E+02 7.31E+02 7.31E+02 7.31E+02 pu240 2.96E+02 2.96E+02 2.96E+02 2.96E+02 2.96E+02 2.96E+02 2.96E+02 pu241 1.76E+02 1.74E+02 1.73E+02 1.72E+02 1.70E+02 1.69E+02 1.67E+02 pu242 6.75E+01 6.75E+01 6.75E+01 6.75E+01 6.75E+01 6.75E+01 6.75E+01 am241 5.09E+00 6.50E+00 7.89E+00 9.28E+00 1.06E+01 1.20E+01 1.34E+01 am242m 1.07E-01 1.07E-01 1.07E-01 1.07E-01 1.07E-01 1.07E-01 1.07E-01 am242 1.40E-02 1.38E-06 1.38E-06 1.38E-06 1.38E-06 1.38E-06 1.37E-06 am243 1.45E+01 1.45E+01 1.45E+01 1.45E+01 1.45E+01 1.45E+01 1.45E+01 cm242 1.77E+00 1.38E+00 1.06E+00 8.20E-01 6.33E-01 4.89E-01 3.77E-01 cm243 4.70E-02 4.68E-02 4.66E-02 4.64E-02 4.62E-02 4.60E-02 4.58E-02 cm244 4.36E+00 4.35E+00 4.32E+00 4.29E+00 4.27E+00 4.24E+00 4.21E+00 cm245 1.55E-01 1.55E-01 1.55E-01 1.55E-01 1.55E-01 1.55E-01 1.55E-01 cm246 2.07E-02 2.07E-02 2.07E-02 2.07E-02 2.07E-02 2.07E-02 2.07E-02 cm247 3.53E-04 3.53E-04 3.53E-04 3.53E-04 3.53E-04 3.53E-04 3.53E-04 cm248 3.16E-05 3.16E-05 3.16E-05 3.16E-05 3.16E-05 3.16E-05 3.16E-05total 1.20E+05 1.20E+05 1.20E+05 1.20E+05 1.20E+05 1.20E+05 1.20E+05

vver440 fuel bu= 36 mwd/ kgufission productsdecay, following reactor irradiation identified by: power=4.96mw, burnup=4464.mwd, flux=4.08E+13n/cm** 2-sec nuclide concentrations, grams basis = single reactor assembly

initial 60.8 d 121.7 d 182.5 d 243.3 d 304.2 d 365.0 d

h 3 6.71E-03 6.64E-03 6.58E-03 6.52E-03 6.46E-03 6.40E-03 6.34E-03 li 6 2.58E-05 2.58E-05 2.58E-05 2.58E-05 2.58E-05 2.58E-05 2.58E-05 li 7 1.37E-06 1.37E-06 1.37E-06 1.37E-06 1.37E-06 1.37E-06 1.37E-06 be 9 2.64E-06 2.64E-06 2.64E-06 2.64E-06 2.64E-06 2.64E-06 2.64E-06 be 10 1.76E-05 1.76E-05 1.76E-05 1.76E-05 1.76E-05 1.76E-05 1.76E-05 c 14 3.57E-06 3.57E-06 3.57E-06 3.57E-06 3.57E-06 3.57E-06 3.57E-06 zn 70 1.09E-06 1.09E-06 1.09E-06 1.09E-06 1.09E-06 1.09E-06 1.09E-06 ga 71 1.06E-05 1.06E-05 1.06E-05 1.06E-05 1.06E-05 1.06E-05 1.06E-05 ge 72 6.85E-04 6.89E-04 6.89E-04 6.89E-04 6.89E-04 6.89E-04 6.89E-04 ge 73 1.97E-03 1.97E-03 1.97E-03 1.97E-03 1.97E-03 1.97E-03 1.97E-03 ge 74 1.72E-03 1.72E-03 1.72E-03 1.72E-03 1.72E-03 1.72E-03 1.72E-03 as 75 1.55E-02 1.55E-02 1.55E-02 1.55E-02 1.55E-02 1.55E-02 1.55E-02 ge 76 4.72E-02 4.72E-02 4.72E-02 4.72E-02 4.72E-02 4.72E-02 4.72E-02 se 76 4.87E-04 4.89E-04 4.89E-04 4.89E-04 4.89E-04 4.89E-04 4.89E-04 se 77 1.06E-01 1.06E-01 1.06E-01 1.06E-01 1.06E-01 1.06E-01 1.06E-01 se 78 3.48E-01 3.48E-01 3.48E-01 3.48E-01 3.48E-01 3.48E-01 3.48E-01 se 79 6.60E-01 6.60E-01 6.60E-01 6.60E-01 6.60E-01 6.60E-01 6.60E-01 br 79 2.12E-06 2.36E-06 2.59E-06 2.82E-06 3.05E-06 3.28E-06 3.51E-06 se 80 1.83E+00 1.83E+00 1.83E+00 1.83E+00 1.83E+00 1.83E+00 1.83E+00 kr 80 7.41E-06 7.42E-06 7.42E-06 7.42E-06 7.42E-06 7.42E-06 7.42E-06 br 81 2.70E+00 2.70E+00 2.70E+00 2.70E+00 2.70E+00 2.70E+00 2.70E+00 se 82 4.40E+00 4.40E+00 4.40E+00 4.40E+00 4.40E+00 4.40E+00 4.40E+00 kr 82 8.52E-02 8.56E-02 8.56E-02 8.56E-02 8.56E-02 8.56E-02 8.56E-02

22

S T U K - Y T O - T R 1 7 0

APPENDIX B NUCLIDE INVENTORY OF SPENT VVER-440 FUEL kr 83 5.61E+00 5.61E+00 5.61E+00 5.61E+00 5.61E+00 5.61E+00 5.61E+00 kr 84 1.56E+01 1.56E+01 1.56E+01 1.56E+01 1.56E+01 1.56E+01 1.56E+01 kr 85 3.09E+00 3.06E+00 3.03E+00 3.00E+00 2.97E+00 2.93E+00 2.90E+00 rb 85 1.26E+01 1.27E+01 1.27E+01 1.27E+01 1.28E+01 1.28E+01 1.28E+01 kr 86 2.52E+01 2.52E+01 2.52E+01 2.52E+01 2.52E+01 2.52E+01 2.52E+01 sr 86 4.80E-02 5.07E-02 5.10E-02 5.10E-02 5.10E-02 5.10E-02 5.10E-02 rb 87 3.28E+01 3.28E+01 3.28E+01 3.28E+01 3.28E+01 3.28E+01 3.28E+01 sr 87 2.48E-04 2.48E-04 2.48E-04 2.48E-04 2.48E-04 2.48E-04 2.48E-04 sr 88 4.70E+01 4.70E+01 4.70E+01 4.70E+01 4.70E+01 4.70E+01 4.70E+01 sr 89 3.99E+00 1.73E+00 7.53E-01 3.27E-01 1.42E-01 6.17E-02 2.68E-02 y 89 5.88E+01 6.11E+01 6.21E+01 6.25E+01 6.27E+01 6.28E+01 6.28E+01 sr 90 7.39E+01 7.36E+01 7.33E+01 7.30E+01 7.27E+01 7.24E+01 7.21E+01 y 90 2.00E-02 1.91E-02 1.90E-02 1.90E-02 1.89E-02 1.88E-02 1.87E-02 zr 90 2.93E+00 3.23E+00 3.53E+00 3.83E+00 4.13E+00 4.43E+00 4.73E+00 y 91 6.19E+00 3.03E+00 1.47E+00 7.17E-01 3.49E-01 1.70E-01 8.25E-02 zr 91 7.52E+01 7.84E+01 8.00E+01 8.07E+01 8.11E+01 8.13E+01 8.14E+01 zr 92 8.68E+01 8.68E+01 8.68E+01 8.68E+01 8.68E+01 8.68E+01 8.68E+01 zr 93 6.39E+01 6.40E+01 6.40E+01 6.40E+01 6.40E+01 6.40E+01 6.40E+01 nb 93 3.18E-06 3.50E-06 3.85E-06 4.23E-06 4.64E-06 5.08E-06 5.56E-06 nb 93m 4.19E-05 4.64E-05 5.09E-05 5.53E-05 5.98E-05 6.41E-05 6.85E-05 zr 94 1.04E+02 1.04E+02 1.04E+02 1.04E+02 1.04E+02 1.04E+02 1.04E+02 nb 94 6.30E-05 6.30E-05 6.30E-05 6.30E-05 6.30E-05 6.30E-05 6.30E-05 zr 95 9.97E+00 5.16E+00 2.67E+00 1.38E+00 7.15E-01 3.70E-01 1.92E-01 nb 95 5.36E+00 4.22E+00 2.62E+00 1.48E+00 8.07E-01 4.29E-01 2.26E-01 nb 95m 6.25E-03 3.42E-03 1.77E-03 9.17E-04 4.74E-04 2.46E-04 1.27E-04 mo 95 8.75E+01 9.35E+01 9.76E+01 1.00E+02 1.01E+02 1.02E+02 1.02E+02 zr 96 1.09E+02 1.09E+02 1.09E+02 1.09E+02 1.09E+02 1.09E+02 1.09E+02 mo 96 5.08E+00 5.08E+00 5.08E+00 5.08E+00 5.08E+00 5.08E+00 5.08E+00 mo 97 1.03E+02 1.03E+02 1.03E+02 1.03E+02 1.03E+02 1.03E+02 1.03E+02

vver440 fuel bu=36 mwd/kgufission productsdecay, following reactor irradiation identified by: power=4.96mw, burnup=4464.mwd, flux=4.08E+13n/cm**2-sec nuclide concentrations, grams basis = single reactor assembly

initial 60.8 d 121.7 d 182.5 d 243.3 d 304.2 d 365.0 d

mo 98 1.12E+02 1.12E+02 1.12E+02 1.12E+02 1.12E+02 1.12E+02 1.12E+02 tc 98 8.46E-04 8.46E-04 8.46E-04 8.46E-04 8.46E-04 8.46E-04 8.46E-04 tc 99 1.06E+02 1.07E+02 1.07E+02 1.07E+02 1.07E+02 1.07E+02 1.07E+02 ru 99 4.21E-03 4.28E-03 4.34E-03 4.40E-03 4.46E-03 4.52E-03 4.57E-03 mo100 1.26E+02 1.26E+02 1.26E+02 1.26E+02 1.26E+02 1.26E+02 1.26E+02 ru100 1.45E+01 1.45E+01 1.45E+01 1.45E+01 1.45E+01 1.45E+01 1.45E+01 ru101 1.03E+02 1.03E+02 1.03E+02 1.03E+02 1.03E+02 1.03E+02 1.03E+02 ru102 1.04E+02 1.04E+02 1.04E+02 1.04E+02 1.04E+02 1.04E+02 1.04E+02 rh102 1.44E-04 1.39E-04 1.33E-04 1.28E-04 1.23E-04 1.18E-04 1.14E-04 ru103 6.72E+00 2.30E+00 7.85E-01 2.68E-01 9.16E-02 3.13E-02 1.07E-02 rh103 5.67E+01 6.11E+01 6.26E+01 6.31E+01 6.33E+01 6.34E+01 6.34E+01 rh103m 6.67E-03 2.28E-03 7.77E-04 2.66E-04 9.07E-05 3.10E-05 1.06E-05 ru104 7.21E+01 7.21E+01 7.21E+01 7.21E+01 7.21E+01 7.21E+01 7.21E+01 pd104 2.92E+01 2.92E+01 2.92E+01 2.92E+01 2.92E+01 2.92E+01 2.92E+01 pd105 5.03E+01 5.05E+01 5.05E+01 5.05E+01 5.05E+01 5.05E+01 5.05E+01 ru106 2.20E+01 1.97E+01 1.76E+01 1.57E+01 1.40E+01 1.25E+01 1.12E+01 rh106 2.32E-05 1.83E-05 1.63E-05 1.45E-05 1.30E-05 1.16E-05 1.03E-05 pd106 2.64E+01 2.88E+01 3.09E+01 3.28E+01 3.45E+01 3.60E+01 3.73E+01 pd107 2.93E+01 2.93E+01 2.93E+01 2.93E+01 2.93E+01 2.93E+01 2.93E+01 ag107 3.25E-06 3.77E-06 4.29E-06 4.81E-06 5.33E-06 5.85E-06 6.37E-06 pd108 1.91E+01 1.91E+01 1.91E+01 1.91E+01 1.91E+01 1.91E+01 1.91E+01 ag108m 2.40E-05 2.40E-05 2.39E-05 2.39E-05 2.39E-05 2.39E-05 2.39E-05 cd108 2.47E-05 2.47E-05 2.47E-05 2.47E-05 2.47E-05 2.47E-05 2.47E-05 ag109 1.16E+01 1.16E+01 1.16E+01 1.16E+01 1.16E+01 1.16E+01 1.16E+01 pd110 5.71E+00 5.71E+00 5.71E+00 5.71E+00 5.71E+00 5.71E+00 5.71E+00 ag110m 1.05E-01 8.85E-02 7.47E-02 6.31E-02 5.33E-02 4.50E-02 3.80E-02 cd110 5.08E+00 5.10E+00 5.11E+00 5.12E+00 5.13E+00 5.14E+00 5.15E+00 cd111 2.92E+00 2.98E+00 2.98E+00 2.98E+00 2.98E+00 2.98E+00 2.98E+00 cd112 1.53E+00 1.53E+00 1.53E+00 1.53E+00 1.53E+00 1.53E+00 1.53E+00 cd113 1.24E-02 1.28E-02 1.28E-02 1.28E-02 1.28E-02 1.28E-02 1.28E-02 cd113m 1.71E-02 1.69E-02 1.68E-02 1.67E-02 1.65E-02 1.64E-02 1.63E-02 in113 8.78E-04 1.02E-03 1.16E-03 1.29E-03 1.43E-03 1.56E-03 1.70E-03 cd114 1.59E+00 1.59E+00 1.59E+00 1.59E+00 1.59E+00 1.59E+00 1.59E+00 sn114 4.62E-05 5.12E-05 5.33E-05 5.43E-05 5.47E-05 5.48E-05 5.49E-05 cd115m 2.08E-03 8.09E-04 3.14E-04 1.22E-04 4.74E-05 1.84E-05 7.16E-06 in115 1.94E-01 1.97E-01 1.98E-01 1.98E-01 1.98E-01 1.98E-01 1.98E-01 sn115 2.34E-02 2.35E-02 2.35E-02 2.35E-02 2.35E-02 2.35E-02 2.35E-02

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23

NUCLIDE INVENTORY OF SPENT VVER-440 FUEL APPENDIX B cd116 6.59E-01 6.59E-01 6.59E-01 6.59E-01 6.59E-01 6.59E-01 6.59E-01 sn116 2.99E-01 2.99E-01 2.99E-01 2.99E-01 2.99E-01 2.99E-01 2.99E-01 sn117 6.16E-01 6.16E-01 6.16E-01 6.16E-01 6.16E-01 6.16E-01 6.16E-01 sn118 4.96E-01 4.96E-01 4.96E-01 4.96E-01 4.96E-01 4.96E-01 4.96E-01 sn119 5.22E-01 5.23E-01 5.23E-01 5.23E-01 5.24E-01 5.24E-01 5.24E-01 sn119m 2.72E-03 2.36E-03 2.04E-03 1.77E-03 1.53E-03 1.33E-03 1.15E-03 sn120 5.12E-01 5.12E-01 5.12E-01 5.12E-01 5.12E-01 5.12E-01 5.12E-01 sn121m 5.69E-03 5.68E-03 5.67E-03 5.65E-03 5.64E-03 5.63E-03 5.62E-03 sb121 5.09E-01 5.11E-01 5.11E-01 5.11E-01 5.11E-01 5.11E-01 5.11E-01 sn122 6.65E-01 6.65E-01 6.65E-01 6.65E-01 6.65E-01 6.65E-01 6.65E-01 te122 3.41E-02 3.45E-02 3.45E-02 3.45E-02 3.45E-02 3.45E-02 3.45E-02 sn123 1.00E-02 7.23E-03 5.22E-03 3.76E-03 2.71E-03 1.96E-03 1.41E-03 sb123 5.96E-01 5.99E-01 6.01E-01 6.03E-01 6.04E-01 6.04E-01 6.05E-01 te123 2.82E-04 3.13E-04 3.36E-04 3.51E-04 3.62E-04 3.70E-04 3.76E-04 te123m 1.07E-04 7.50E-05 5.28E-05 3.71E-05 2.61E-05 1.83E-05 1.29E-05

vver440 fuel bu=36 mwd/kgu fission productsdecay, following reactor irradiation identified by: power=4.96mw, burnup=4464.mwd, flux=4.08E+13n/cm**2-sec nuclide concentrations, grams basis = single reactor assembly

initial 60.8 d 121.7 d 182.5 d 243.3 d 304.2 d 365.0 d

sn124 1.11E+00 1.11E+00 1.11E+00 1.11E+00 1.11E+00 1.11E+00 1.11E+00 sb124 5.97E-03 2.96E-03 1.47E-03 7.30E-04 3.62E-04 1.80E-04 8.92E-05 te124 2.58E-02 2.88E-02 3.03E-02 3.10E-02 3.14E-02 3.16E-02 3.17E-02 sb125 1.01E+00 9.74E-01 9.34E-01 8.95E-01 8.58E-01 8.23E-01 7.89E-01 te125 3.38E-01 3.79E-01 4.20E-01 4.59E-01 4.96E-01 5.32E-01 5.67E-01 te125m 1.24E-02 1.29E-02 1.28E-02 1.25E-02 1.21E-02 1.16E-02 1.12E-02 sn126 2.55E+00 2.55E+00 2.55E+00 2.55E+00 2.55E+00 2.55E+00 2.55E+00 te126 4.63E-02 4.70E-02 4.70E-02 4.70E-02 4.70E-02 4.70E-02 4.70E-02 te127 4.54E-03 5.02E-04 3.41E-04 2.32E-04 1.57E-04 1.07E-04 7.26E-05 te127m 2.03E-01 1.44E-01 9.75E-02 6.62E-02 4.50E-02 3.05E-02 2.07E-02 i127 5.46E+00 5.57E+00 5.61E+00 5.64E+00 5.66E+00 5.68E+00 5.69E+00 te128 1.20E+01 1.20E+01 1.20E+01 1.20E+01 1.20E+01 1.20E+01 1.20E+01 xe128 3.38E-01 3.38E-01 3.38E-01 3.38E-01 3.38E-01 3.38E-01 3.38E-01 te129m 2.79E-01 8.00E-02 2.28E-02 6.50E-03 1.85E-03 5.28E-04 1.51E-04 i129 2.39E+01 2.42E+01 2.42E+01 2.42E+01 2.42E+01 2.42E+01 2.42E+01 xe129 1.96E-03 2.01E-03 2.01E-03 2.01E-03 2.01E-03 2.01E-03 2.01E-03 te130 4.89E+01 4.89E+01 4.89E+01 4.89E+01 4.89E+01 4.89E+01 4.89E+01 xe130 9.52E-01 9.54E-01 9.54E-01 9.54E-01 9.54E-01 9.54E-01 9.54E-01 xe131 5.47E+01 5.59E+01 5.59E+01 5.59E+01 5.59E+01 5.59E+01 5.59E+01 xe132 1.48E+02 1.49E+02 1.49E+02 1.49E+02 1.49E+02 1.49E+02 1.49E+02 ba132 2.79E-05 2.86E-05 2.86E-05 2.86E-05 2.86E-05 2.86E-05 2.86E-05 cs133 1.53E+02 1.55E+02 1.55E+02 1.55E+02 1.55E+02 1.55E+02 1.55E+02 xe134 2.05E+02 2.05E+02 2.05E+02 2.05E+02 2.05E+02 2.05E+02 2.05E+02 cs134 1.62E+01 1.53E+01 1.45E+01 1.37E+01 1.30E+01 1.23E+01 1.16E+01 ba134 5.36E+00 6.24E+00 7.08E+00 7.87E+00 8.61E+00 9.32E+00 9.99E+00 cs135 4.77E+01 4.78E+01 4.78E+01 4.78E+01 4.78E+01 4.78E+01 4.78E+01 ba135 4.07E-02 4.07E-02 4.07E-02 4.07E-02 4.07E-02 4.07E-02 4.07E-02 xe136 3.09E+02 3.09E+02 3.09E+02 3.09E+02 3.09E+02 3.09E+02 3.09E+02 ba136 2.27E+00 2.36E+00 2.36E+00 2.36E+00 2.36E+00 2.36E+00 2.36E+00 cs137 1.67E+02 1.67E+02 1.66E+02 1.65E+02 1.65E+02 1.64E+02 1.64E+02 ba137 5.56E+00 6.21E+00 6.85E+00 7.49E+00 8.12E+00 8.75E+00 9.38E+00 ba137m 2.58E-05 2.55E-05 2.54E-05 2.53E-05 2.52E-05 2.51E-05 2.50E-05 ba138 1.74E+02 1.74E+02 1.74E+02 1.74E+02 1.74E+02 1.74E+02 1.74E+02 la138 1.15E-03 1.15E-03 1.15E-03 1.15E-03 1.15E-03 1.15E-03 1.15E-03 la139 1.64E+02 1.64E+02 1.64E+02 1.64E+02 1.64E+02 1.64E+02 1.64E+02 ce140 1.73E+02 1.76E+02 1.76E+02 1.76E+02 1.76E+02 1.76E+02 1.76E+02 ce141 7.63E+00 2.10E+00 5.73E-01 1.57E-01 4.28E-02 1.17E-02 3.19E-03 pr141 1.43E+02 1.49E+02 1.50E+02 1.51E+02 1.51E+02 1.51E+02 1.51E+02 ce142 1.53E+02 1.53E+02 1.53E+02 1.53E+02 1.53E+02 1.53E+02 1.53E+02 nd142 2.58E+00 2.58E+00 2.58E+00 2.58E+00 2.58E+00 2.58E+00 2.58E+00 nd143 1.08E+02 1.11E+02 1.11E+02 1.11E+02 1.11E+02 1.11E+02 1.11E+02 ce144 4.87E+01 4.20E+01 3.63E+01 3.13E+01 2.70E+01 2.33E+01 2.01E+01 pr144 2.07E-03 1.77E-03 1.53E-03 1.32E-03 1.14E-03 9.80E-04 8.45E-04 pr144m 1.20E-05 1.03E-05 8.91E-06 7.68E-06 6.63E-06 5.71E-06 4.93E-06 nd144 1.27E+02 1.34E+02 1.40E+02 1.45E+02 1.49E+02 1.53E+02 1.56E+02 nd145 9.07E+01 9.07E+01 9.07E+01 9.07E+01 9.07E+01 9.07E+01 9.07E+01 pm145 2.88E-06 2.88E-06 2.89E-06 2.89E-06 2.88E-06 2.88E-06 2.87E-06 nd146 9.35E+01 9.35E+01 9.35E+01 9.35E+01 9.35E+01 9.35E+01 9.35E+01 pm146 1.02E-03 9.97E-04 9.76E-04 9.56E-04 9.36E-04 9.17E-04 8.98E-04 sm146 7.40E-04 7.47E-04 7.54E-04 7.61E-04 7.67E-04 7.74E-04 7.80E-04 pm147 2.33E+01 2.33E+01 2.23E+01 2.14E+01 2.05E+01 1.96E+01 1.87E+01 sm147 9.27E+00 1.03E+01 1.13E+01 1.23E+01 1.32E+01 1.41E+01 1.49E+01

24

S T U K - Y T O - T R 1 7 0

APPENDIX B NUCLIDE INVENTORY OF SPENT VVER-440 FUELvver440 fuel bu= 36 mwd/ kgu fission productsdecay, following reactor irradiation identified by: power=4.96mw, burnup=4464.mwd, flux=4.08E+13n/cm**2-sec nuclide concentrations, grams basis = single reactor assembly

initial 60.8 d 121.7 d 182.5 d 243.3 d 304.2 d 365.0 d

nd148 4.94E+01 4.94E+01 4.94E+01 4.94E+01 4.94E+01 4.94E+01 4.94E+01 pm148 1.46E-01 4.78E-04 1.52E-04 5.47E-05 1.97E-05 7.09E-06 2.56E-06 pm148m 1.70E-01 6.12E-02 2.20E-02 7.94E-03 2.86E-03 1.03E-03 3.71E-04 sm148 1.56E+01 1.59E+01 1.59E+01 1.59E+01 1.59E+01 1.59E+01 1.59E+01 sm149 3.31E-01 5.45E-01 5.45E-01 5.45E-01 5.45E-01 5.45E-01 5.45E-01 nd150 2.38E+01 2.38E+01 2.38E+01 2.38E+01 2.38E+01 2.38E+01 2.38E+01 sm150 4.10E+01 4.10E+01 4.10E+01 4.10E+01 4.10E+01 4.10E+01 4.10E+01 sm151 2.10E+00 2.13E+00 2.13E+00 2.13E+00 2.13E+00 2.12E+00 2.12E+00 eu151 2.04E-03 4.78E-03 7.52E-03 1.02E-02 1.30E-02 1.57E-02 1.84E-02 sm152 1.70E+01 1.70E+01 1.70E+01 1.70E+01 1.70E+01 1.70E+01 1.70E+01 eu152 5.01E-03 4.97E-03 4.93E-03 4.89E-03 4.84E-03 4.80E-03 4.76E-03 gd152 7.12E-03 7.14E-03 7.15E-03 7.16E-03 7.17E-03 7.19E-03 7.20E-03 eu153 1.58E+01 1.60E+01 1.60E+01 1.60E+01 1.60E+01 1.60E+01 1.60E+01 gd153 5.75E-04 4.83E-04 4.06E-04 3.41E-04 2.86E-04 2.40E-04 2.02E-04 sm154 4.90E+00 4.90E+00 4.90E+00 4.90E+00 4.90E+00 4.90E+00 4.90E+00 eu154 3.18E+00 3.14E+00 3.10E+00 3.05E+00 3.01E+00 2.97E+00 2.93E+00 gd154 2.31E-01 2.73E-01 3.15E-01 3.56E-01 3.97E-01 4.37E-01 4.77E-01 eu155 7.23E-01 7.06E-01 6.89E-01 6.72E-01 6.55E-01 6.39E-01 6.24E-01 gd155 5.35E-03 2.30E-02 4.02E-02 5.69E-02 7.33E-02 8.93E-02 1.05E-01 gd156 9.49E+00 1.01E+01 1.01E+01 1.01E+01 1.01E+01 1.01E+01 1.01E+01 gd157 1.39E-02 1.69E-02 1.69E-02 1.69E-02 1.69E-02 1.69E-02 1.69E-02 gd158 2.39E+00 2.39E+00 2.39E+00 2.39E+00 2.39E+00 2.39E+00 2.39E+00 tb159 3.02E-01 3.03E-01 3.03E-01 3.03E-01 3.03E-01 3.03E-01 3.03E-01 gd160 1.33E-01 1.33E-01 1.33E-01 1.33E-01 1.33E-01 1.33E-01 1.33E-01 tb160 1.11E-02 6.20E-03 3.46E-03 1.93E-03 1.08E-03 6.02E-04 3.36E-04 dy160 2.33E-02 2.82E-02 3.09E-02 3.24E-02 3.33E-02 3.38E-02 3.40E-02 dy161 4.46E-02 4.60E-02 4.60E-02 4.60E-02 4.60E-02 4.60E-02 4.60E-02 dy162 3.48E-02 3.48E-02 3.48E-02 3.48E-02 3.48E-02 3.48E-02 3.48E-02 dy163 2.77E-02 2.77E-02 2.77E-02 2.77E-02 2.77E-02 2.77E-02 2.77E-02 dy164 6.28E-03 6.28E-03 6.28E-03 6.28E-03 6.28E-03 6.28E-03 6.28E-03 ho165 8.71E-03 8.72E-03 8.72E-03 8.72E-03 8.72E-03 8.72E-03 8.72E-03 ho166m 2.93E-05 2.93E-05 2.93E-05 2.93E-05 2.93E-05 2.93E-05 2.93E-05 er166 1.88E-03 1.89E-03 1.89E-03 1.89E-03 1.89E-03 1.89E-03 1.89E-03 er167 3.58E-05 3.58E-05 3.58E-05 3.58E-05 3.58E-05 3.58E-05 3.58E-05 er168 4.63E-05 4.63E-05 4.63E-05 4.63E-05 4.63E-05 4.63E-05 4.63E-05total 4.59E+03 4.59E+03 4.59E+03 4.59E+03 4.59E+03 4.59E+03 4.59E+03

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NEUTRON SOURCE OF SPENT BWR FUEL APPENDIX Cneutron source intensity as a function of timebu=37816, bwr, 5 cyc, e=2.746, vf=0.426, 8x8-e2-275-1-200-10alpha-n neutron source, neutrons/sec/basis basis = single reactor assembly

initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d

pb210 6.12E-16 6.24E-16 6.14E-16 6.60E-16 8.02E-16 1.09E-15 1.59E-15 bi210m 1.37E-16 1.37E-16 1.37E-16 1.37E-16 1.37E-16 1.37E-16 1.37E-16 bi210 1.57E-13 1.59E-13 1.57E-13 1.68E-13 2.05E-13 2.79E-13 4.07E-13 bi211 4.32E-05 6.67E-05 9.65E-05 1.33E-04 1.76E-04 2.25E-04 2.79E-04 bi212 1.38E-01 5.10E-01 9.21E-01 1.24E+00 1.45E+00 1.57E+00 1.64E+00 bi213 7.53E-07 1.26E-07 1.34E-07 1.44E-07 1.57E-07 1.73E-07 1.92E-07 bi214 6.99E-12 2.91E-11 9.20E-11 2.20E-10 4.38E-10 7.68E-10 1.24E-09 po210 1.23E-07 1.95E-07 1.91E-07 2.01E-07 2.37E-07 3.15E-07 4.54E-07 po211 1.71E-07 2.64E-07 3.82E-07 5.27E-07 6.97E-07 8.90E-07 1.10E-06 po212 7.07E-01 2.61E+00 4.72E+00 6.34E+00 7.41E+00 8.05E+00 8.38E+00 po213 9.93E-05 1.67E-05 1.77E-05 1.90E-05 2.08E-05 2.29E-05 2.53E-05 po214 1.91E-05 2.60E-07 8.19E-07 1.96E-06 3.90E-06 6.85E-06 1.10E-05 po215 6.10E-05 9.42E-05 1.36E-04 1.88E-04 2.49E-04 3.18E-04 3.94E-04 po216 5.51E-01 2.04E+00 3.68E+00 4.94E+00 5.78E+00 6.28E+00 6.54E+00 po218 2.96E-08 1.23E-07 3.90E-07 9.33E-07 1.85E-06 3.26E-06 5.24E-06 at217 6.44E-05 1.08E-05 1.15E-05 1.23E-05 1.35E-05 1.48E-05 1.64E-05 rn218 1.54E-05 2.21E-17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 rn219 4.85E-05 7.49E-05 1.08E-04 1.50E-04 1.98E-04 2.53E-04 3.13E-04 rn220 4.37E-01 1.62E+00 2.92E+00 3.92E+00 4.58E+00 4.97E+00 5.18E+00 rn222 2.16E-08 9.02E-08 2.85E-07 6.81E-07 1.35E-06 2.38E-06 3.82E-06 fr221 4.70E-05 7.88E-06 8.36E-06 9.00E-06 9.82E-06 1.08E-05 1.20E-05 fr223 1.88E-11 2.84E-11 4.10E-11 5.66E-11 7.48E-11 9.56E-11 1.19E-10 ra222 1.19E-05 1.71E-17 3.58E-29 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ra223 2.81E-05 4.34E-05 6.28E-05 8.66E-05 1.15E-04 1.46E-04 1.81E-04 ra224 3.09E-01 1.14E+00 2.06E+00 2.77E+00 3.24E+00 3.52E+00 3.66E+00 ra226 1.27E-08 5.27E-08 1.66E-07 3.98E-07 7.92E-07 1.39E-06 2.24E-06 ac225 3.38E-05 5.67E-06 6.01E-06 6.47E-06 7.06E-06 7.77E-06 8.60E-06 ac227 2.11E-07 3.19E-07 4.62E-07 6.37E-07 8.42E-07 1.08E-06 1.33E-06 ac228 2.07E-17 5.00E-17 9.04E-17 1.39E-16 1.95E-16 2.55E-16 3.19E-16 th226 1.08E-05 1.54E-17 3.23E-29 0.00E+00 0.00E+00 0.00E+00 0.00E+00 th227 3.13E-05 4.79E-05 6.93E-05 9.55E-05 1.26E-04 1.61E-04 2.00E-04 th228 2.59E-01 9.63E-01 1.74E+00 2.33E+00 2.73E+00 2.96E+00 3.08E+00 th229 3.18E-06 3.31E-06 3.51E-06 3.78E-06 4.12E-06 4.54E-06 5.03E-06 th230 1.73E-05 6.60E-05 1.57E-04 2.91E-04 4.66E-04 6.81E-04 9.36E-04 th232 1.33E-09 2.31E-09 3.28E-09 4.25E-09 5.22E-09 6.19E-09 7.16E-09 pa231 1.17E-04 1.66E-04 2.15E-04 2.64E-04 3.13E-04 3.62E-04 4.10E-04 u230 8.48E-06 1.21E-17 2.54E-29 0.00E+00 0.00E+00 0.00E+00 0.00E+00 u231 1.05E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 u232 7.71E-01 1.83E+00 2.42E+00 2.74E+00 2.89E+00 2.96E+00 2.97E+00 u233 4.03E-04 7.20E-04 1.04E-03 1.35E-03 1.67E-03 1.99E-03 2.32E-03 u234 1.45E+00 3.66E+00 5.87E+00 8.05E+00 1.02E+01 1.23E+01 1.44E+01 u235 6.26E-01 6.26E-01 6.26E-01 6.26E-01 6.26E-01 6.26E-01 6.26E-01 u236 1.55E+01 1.55E+01 1.55E+01 1.55E+01 1.55E+01 1.55E+01 1.55E+01 u238 1.51E+01 1.51E+01 1.51E+01 1.51E+01 1.51E+01 1.51E+01 1.51E+01 np235 2.87E-05 6.86E-06 1.64E-06 3.91E-07 9.35E-08 2.23E-08 5.33E-09 np237 2.75E+01 2.79E+01 2.80E+01 2.81E+01 2.82E+01 2.83E+01 2.84E+01 pu236 8.44E+01 4.96E+01 2.90E+01 1.70E+01 9.94E+00 5.82E+00 3.41E+00 pu237 3.45E-02 1.21E-07 4.27E-13 1.51E-18 5.30E-24 1.87E-29 0.00E+00

neutron source intensity as a function of timebu=37816, bwr, 5 cyc, e=2.746, vf=0.426, 8x8-e2-275-1-200-10alpha-n neutron source, neutrons/sec/basis basis = single reactor assembly

initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d

pu238 5.66E+05 6.08E+05 5.99E+05 5.89E+05 5.79E+05 5.68E+05 5.58E+05 pu239 3.84E+04 3.89E+04 3.89E+04 3.89E+04 3.89E+04 3.89E+04 3.89E+04 pu240 6.94E+04 6.96E+04 6.97E+04 6.98E+04 6.99E+04 7.00E+04 7.01E+04 pu241 3.40E+02 3.05E+02 2.74E+02 2.46E+02 2.21E+02 1.98E+02 1.78E+02 pu242 2.84E+02 2.84E+02 2.84E+02 2.84E+02 2.84E+02 2.84E+02 2.84E+02 pu244 1.53E-12 8.95E-12 1.64E-11 2.38E-11 3.12E-11 3.87E-11 4.61E-11 am239 8.45E-05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 am240 1.29E-04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 am241 3.25E+04 1.07E+05 1.74E+05 2.34E+05 2.87E+05 3.35E+05 3.77E+05 am242m 7.74E+00 7.66E+00 7.57E+00 7.49E+00 7.41E+00 7.33E+00 7.25E+00 am243 4.88E+03 4.88E+03 4.88E+03 4.88E+03 4.88E+03 4.87E+03 4.87E+03 cm241 4.19E-02 1.30E-09 4.00E-17 1.24E-24 0.00E+00 0.00E+00 0.00E+00 cm242 1.40E+07 4.35E+05 1.57E+04 2.74E+03 2.31E+03 2.28E+03 2.25E+03

26

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APPENDIX C NEUTRON SOURCE OF SPENT BWR FUEL cm243 6.39E+03 6.05E+03 5.73E+03 5.43E+03 5.14E+03 4.87E+03 4.61E+03 cm244 1.09E+06 9.98E+05 9.16E+05 8.40E+05 7.71E+05 7.08E+05 6.50E+05 cm245 7.04E+01 7.03E+01 7.03E+01 7.03E+01 7.03E+01 7.03E+01 7.03E+01 cm246 2.28E+01 2.28E+01 2.28E+01 2.28E+01 2.27E+01 2.27E+01 2.27E+01 cm247 1.04E-04 1.04E-04 1.04E-04 1.04E-04 1.04E-04 1.04E-04 1.04E-04 cm248 5.74E-04 5.74E-04 5.74E-04 5.74E-04 5.74E-04 5.74E-04 5.74E-04 cm250 2.94E-11 2.94E-11 2.94E-11 2.94E-11 2.94E-11 2.94E-11 2.94E-11 bk249 7.29E-05 1.24E-05 2.11E-06 3.58E-07 6.08E-08 1.03E-08 1.76E-09 cf249 2.97E-03 1.74E-02 1.98E-02 2.01E-02 2.01E-02 2.00E-02 1.99E-02 cf250 8.67E-02 7.73E-02 6.86E-02 6.10E-02 5.41E-02 4.81E-02 4.27E-02 cf251 7.31E-04 7.29E-04 7.28E-04 7.27E-04 7.25E-04 7.24E-04 7.23E-04 cf252 2.18E-01 1.21E-01 6.74E-02 3.75E-02 2.08E-02 1.16E-02 6.44E-03 cf253 5.97E-05 8.84E-19 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 cf254 1.46E-08 1.24E-12 1.05E-16 8.91E-21 7.55E-25 6.40E-29 0.00E+00 es253 2.10E-02 1.79E-13 1.68E-25 0.00E+00 0.00E+00 0.00E+00 0.00E+00 es254 6.61E-05 8.45E-06 1.08E-06 1.38E-07 1.76E-08 2.25E-09 2.88E-10 es255 1.06E-06 5.14E-13 2.48E-19 1.20E-25 0.00E+00 0.00E+00 0.00E+00total 1.58E+07 2.27E+06 1.82E+06 1.79E+06 1.76E+06 1.73E+06 1.71E+06

neutron source intensity as a function of timebu=37816, bwr, 5 cyc, e=2.746, vf=0.426, 8x8-e2-275-1-200-10spontaneous fission neutron source, neutrons/sec/basis basis = single reactor assembly

initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d

th230 4.43E-10 1.69E-09 4.04E-09 7.46E-09 1.19E-08 1.75E-08 2.40E-08 pa231 1.47E-08 2.08E-08 2.70E-08 3.31E-08 3.92E-08 4.53E-08 5.15E-08 u232 4.74E-05 1.13E-04 1.49E-04 1.68E-04 1.78E-04 1.82E-04 1.82E-04 u234 3.13E-03 7.86E-03 1.26E-02 1.73E-02 2.19E-02 2.64E-02 3.09E-02 u235 7.65E-03 7.66E-03 7.66E-03 7.66E-03 7.66E-03 7.66E-03 7.66E-03 u236 2.32E+00 2.32E+00 2.32E+00 2.32E+00 2.32E+00 2.32E+00 2.33E+00 u237 2.02E-06 1.12E-11 1.00E-11 9.00E-12 8.08E-12 7.25E-12 6.51E-12 u238 2.12E+03 2.12E+03 2.12E+03 2.12E+03 2.12E+03 2.12E+03 2.12E+03 u239 1.01E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 np236 2.52E-08 2.52E-08 2.52E-08 2.52E-08 2.52E-08 2.52E-08 2.52E-08 np238 1.16E-05 2.25E-12 2.22E-12 2.20E-12 2.17E-12 2.15E-12 2.13E-12 np239 2.06E-02 5.06E-08 5.06E-08 5.05E-08 5.05E-08 5.05E-08 5.05E-08 pu236 5.71E+00 3.36E+00 1.96E+00 1.15E+00 6.73E-01 3.94E-01 2.31E-01 pu238 1.05E+05 1.12E+05 1.11E+05 1.09E+05 1.07E+05 1.05E+05 1.03E+05 pu239 2.06E+01 2.08E+01 2.08E+01 2.08E+01 2.08E+01 2.08E+01 2.08E+01 pu240 4.62E+05 4.62E+05 4.63E+05 4.64E+05 4.65E+05 4.65E+05 4.66E+05 pu241 1.22E+01 1.10E+01 9.86E+00 8.85E+00 7.94E+00 7.13E+00 6.39E+00 pu242 2.21E+05 2.21E+05 2.21E+05 2.21E+05 2.21E+05 2.21E+05 2.21E+05 pu243 1.86E-03 4.48E-15 4.48E-15 4.48E-15 4.48E-15 4.48E-15 4.48E-15 pu244 3.66E-07 2.14E-06 3.92E-06 5.69E-06 7.47E-06 9.25E-06 1.10E-05 am241 1.25E+01 4.13E+01 6.71E+01 9.01E+01 1.11E+02 1.29E+02 1.45E+02 am242m 3.68E+01 3.64E+01 3.60E+01 3.56E+01 3.52E+01 3.48E+01 3.44E+01 am242 2.97E+02 3.95E-02 3.91E-02 3.87E-02 3.82E-02 3.78E-02 3.74E-02 am243 2.24E+01 2.25E+01 2.25E+01 2.25E+01 2.24E+01 2.24E+01 2.24E+01 am244 4.18E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 cm242 6.98E+07 2.17E+06 7.82E+04 1.37E+04 1.15E+04 1.14E+04 1.12E+04 cm243 1.39E+02 1.31E+02 1.24E+02 1.18E+02 1.12E+02 1.06E+02 1.00E+02 cm244 1.42E+08 1.30E+08 1.19E+08 1.10E+08 1.01E+08 9.23E+07 8.48E+07 cm245 1.91E+01 1.91E+01 1.91E+01 1.91E+01 1.91E+01 1.91E+01 1.91E+01 cm246 8.21E+05 8.21E+05 8.20E+05 8.20E+05 8.20E+05 8.20E+05 8.19E+05 cm248 9.39E+03 9.40E+03 9.40E+03 9.40E+03 9.40E+03 9.40E+03 9.40E+03 cm250 1.25E-02 1.25E-02 1.25E-02 1.25E-02 1.25E-02 1.25E-02 1.25E-02 bk249 4.02E-01 6.84E-02 1.16E-02 1.97E-03 3.35E-04 5.70E-05 9.68E-06 cf249 1.80E-03 1.06E-02 1.20E-02 1.22E-02 1.22E-02 1.22E-02 1.21E-02 cf250 7.26E+03 6.47E+03 5.75E+03 5.11E+03 4.53E+03 4.03E+03 3.58E+03 cf252 7.64E+05 4.24E+05 2.36E+05 1.31E+05 7.29E+04 4.05E+04 2.25E+04 cf254 6.28E+02 5.32E-02 4.51E-06 3.83E-10 3.24E-14 2.75E-18 2.33E-22 es253 1.56E-01 1.33E-12 1.25E-24 0.00E+00 0.00E+00 0.00E+00 0.00E+00 es255 3.06E-03 1.48E-09 7.15E-16 3.45E-22 1.66E-28 0.00E+00 0.00E+00total 2.14E+08 1.34E+08 1.21E+08 1.11E+08 1.02E+08 9.40E+07 8.64E+07--------------------------------------------------------------------------------------------total 2.30E+08 1.37E+08 1.23E+08 1.13E+08 1.04E+08 9.58E+07 8.81E+07

S T U K - Y T O - T R 1 7 0

27

NEUTRON SOURCE OF SPENT BWR FUEL APPENDIX Calpha-n neutron source spectrum as a function of time(using reaction spectra for uranium dioxide)bu=37816, bwr, 5 cyc, e=2.746, vf=0.426, 8x8-e2-275-1-200-10alpha-n neutron spectra, neutrons/sec/basis basis = single reactor assembly

boundaries, mev initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d

1 6.43E+00 - 2.00E+01 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2 3.00E+00 - 6.43E+00 6.765E+06 6.296E+05 4.227E+05 4.086E+05 4.007E+05 3.930E+05 3.855E+05 3 1.85E+00 - 3.00E+00 7.498E+06 1.178E+06 9.720E+05 9.528E+05 9.394E+05 9.261E+05 9.129E+05 4 1.40E+00 - 1.85E+00 1.049E+06 2.728E+05 2.492E+05 2.455E+05 2.423E+05 2.392E+05 2.361E+05 5 9.00E-01 - 1.40E+00 3.829E+05 1.423E+05 1.357E+05 1.340E+05 1.324E+05 1.308E+05 1.293E+05 6 4.00E-01 - 9.00E-01 7.087E+04 3.946E+04 3.887E+04 3.845E+04 3.800E+04 3.756E+04 3.712E+04 7 1.00E-01 - 4.00E-01 1.305E+04 6.656E+03 6.506E+03 6.421E+03 6.334E+03 6.247E+03 6.161E+03 8 1.70E-02 - 1.00E-01 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 9 3.00E-03 - 1.70E-02 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0010 5.50E-04 - 3.00E-03 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0011 1.00E-04 - 5.50E-04 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0012 3.00E-05 - 1.00E-04 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0013 1.00E-05 - 3.00E-05 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0014 3.05E-06 - 1.00E-05 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0015 1.77E-06 - 3.05E-06 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0016 1.30E-06 - 1.77E-06 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0017 1.13E-06 - 1.30E-06 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0018 1.00E-06 - 1.13E-06 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0019 8.00E-07 - 1.00E-06 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0020 4.00E-07 - 8.00E-07 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0021 3.25E-07 - 4.00E-07 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0022 2.25E-07 - 3.25E-07 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0023 1.00E-07 - 2.25E-07 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0024 5.00E-08 - 1.00E-07 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0025 3.00E-08 - 5.00E-08 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0026 1.00E-08 - 3.00E-08 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0027 1.00E-11 - 1.00E-08 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00

1.578E+07 2.269E+06 1.825E+06 1.786E+06 1.759E+06 1.733E+06 1.707E+06

spontaneous fission neutron source spectrum as a function of timebu=37816, bwr, 5 cyc, e=2.746, vf=0.426, 8x8-e2-275-1-200-10spontaneous fission neutron spectra, neutrons/ sec/ basis basis = single reactor assembly



2.137E+08 1.344E+08 1.214E+08 1.114E+08 1.023E+08 9.402E+07 8.641E+07

28

S T U K - Y T O - T R 1 7 0

APPENDIX C NEUTRON SOURCE OF SPENT BWR FUELtotal (alpha-n plus spon. fission) neutron source spectrum as a function of time(using reaction spectra for uranium dioxide)bu=37816, bwr, 5 cyc, e=2.746, vf=0.426, 8x8-e2-275-1-200-10neutron spectra, neutrons/sec/basis basis = single reactor assembly



2.295E+08 1.367E+08 1.232E+08 1.132E+08 1.041E+08 9.576E+07 8.812E+07

S T U K - Y T O - T R 1 7 0

29

GAMMA SOURCE OF SPENT BWR FUEL APPENDIX Dprincipal photon sources in group 1, photons/secmean energy = 0.0100 mev. nuclides exceeding 1.0E-02 of total group release rate (3.08E+14) at4091.7 dnuclide time after discharge

initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d kr 85 7.78E+12 6.73E+12 5.83E+12 5.04E+12 4.36E+12 3.77E+12 3.26E+12 sr 90 5.30E+13 5.01E+13 4.75E+13 4.49E+13 4.25E+13 4.02E+13 3.81E+13 y 90 2.70E+14 2.46E+14 2.33E+14 2.21E+14 2.09E+14 1.98E+14 1.87E+14 cs137 7.09E+13 6.73E+13 6.39E+13 6.07E+13 5.76E+13 5.47E+13 5.20E+13 eu154 8.77E+12 7.32E+12 6.11E+12 5.10E+12 4.26E+12 3.55E+12 2.97E+12

principal photon sources in group 2, photons/secmean energy = 0.0300 mev. nuclides exceeding 1.0E-02 of total group release rate (1.41E+14) at4091.7 dnuclide time after discharge

initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d sr 90 1.50E+13 1.42E+13 1.34E+13 1.27E+13 1.20E+13 1.14E+13 1.08E+13 y 90 8.79E+13 8.03E+13 7.59E+13 7.19E+13 6.80E+13 6.44E+13 6.09E+13 sb125 2.52E+13 1.43E+13 8.10E+12 4.59E+12 2.60E+12 1.47E+12 8.32E+11 cs137 1.98E+13 1.88E+13 1.78E+13 1.69E+13 1.61E+13 1.53E+13 1.45E+13 ba137m 5.75E+13 5.42E+13 5.15E+13 4.89E+13 4.64E+13 4.41E+13 4.19E+13


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d kr 85 1.39E+12 1.20E+12 1.04E+12 9.01E+11 7.79E+11 6.74E+11 5.83E+11 sr 90 8.85E+12 8.38E+12 7.93E+12 7.50E+12 7.10E+12 6.72E+12 6.36E+12 y 90 6.08E+13 5.55E+13 5.25E+13 4.97E+13 4.70E+13 4.45E+13 4.21E+13 cs137 1.15E+13 1.09E+13 1.04E+13 9.84E+12 9.35E+12 8.88E+12 8.43E+12 eu154 1.13E+13 9.40E+12 7.84E+12 6.55E+12 5.46E+12 4.56E+12 3.81E+12 eu155 4.37E+12 3.13E+12 2.25E+12 1.61E+12 1.16E+12 8.31E+11 5.96E+11


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d sr 90 4.22E+12 3.99E+12 3.78E+12 3.58E+12 3.38E+12 3.20E+12 3.03E+12 y 90 3.52E+13 3.21E+13 3.04E+13 2.88E+13 2.72E+13 2.57E+13 2.44E+13 cs137 5.38E+12 5.11E+12 4.86E+12 4.61E+12 4.38E+12 4.16E+12 3.95E+12 eu155 6.63E+12 4.76E+12 3.41E+12 2.45E+12 1.76E+12 1.26E+12 9.05E+11


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d sr 90 2.40E+12 2.27E+12 2.15E+12 2.04E+12 1.93E+12 1.82E+12 1.72E+12 y 90 2.47E+13 2.26E+13 2.14E+13 2.02E+13 1.91E+13 1.81E+13 1.71E+13 cs137 3.02E+12 2.87E+12 2.72E+12 2.58E+12 2.45E+12 2.33E+12 2.21E+12 eu154 2.09E+13 1.75E+13 1.46E+13 1.22E+13 1.02E+13 8.48E+12 7.08E+12 eu155 3.78E+12 2.72E+12 1.95E+12 1.40E+12 1.00E+12 7.20E+11 5.17E+11


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d sr 90 1.71E+12 1.62E+12 1.53E+12 1.45E+12 1.37E+12 1.30E+12 1.23E+12 y 90 2.55E+13 2.33E+13 2.20E+13 2.09E+13 1.97E+13 1.87E+13 1.77E+13 cs137 2.13E+12 2.02E+12 1.92E+12 1.82E+12 1.73E+12 1.65E+12 1.56E+12


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d sr 90 7.77E+11 7.35E+11 6.96E+11 6.58E+11 6.23E+11 5.89E+11 5.58E+11 y 90 2.83E+13 2.59E+13 2.45E+13 2.32E+13 2.19E+13 2.07E+13 1.96E+13 cs137 1.08E+12 1.02E+12 9.73E+11 9.23E+11 8.77E+11 8.33E+11 7.91E+11 eu154 3.20E+12 2.67E+12 2.23E+12 1.86E+12 1.55E+12 1.30E+12 1.08E+12

30

S T U K - Y T O - T R 1 7 0

APPENDIX D GAMMA SOURCE OF SPENT BWR FUELprincipal photon sources in group 8, photons/secmean energy = 0.6500 mev. nuclides exceeding 1.0E-02 of total group release rate (5.76E+14) at4091.7 dnuclide time after discharge

initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d y 90 1.20E+13 1.09E+13 1.03E+13 9.79E+12 9.26E+12 8.76E+12 8.29E+12 cs134 2.49E+15 1.17E+15 5.52E+14 2.60E+14 1.22E+14 5.76E+13 2.71E+13 ba137m 6.47E+14 6.10E+14 5.79E+14 5.50E+14 5.22E+14 4.96E+14 4.71E+14 eu154 2.68E+13 2.24E+13 1.87E+13 1.56E+13 1.30E+13 1.09E+13 9.06E+12


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d y 90 1.56E+12 1.43E+12 1.35E+12 1.28E+12 1.21E+12 1.14E+12 1.08E+12 cs134 3.05E+13 1.43E+13 6.75E+12 3.18E+12 1.50E+12 7.05E+11 3.32E+11 eu154 3.32E+13 2.77E+13 2.31E+13 1.93E+13 1.61E+13 1.35E+13 1.12E+13


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d y 90 1.99E+11 1.82E+11 1.72E+11 1.63E+11 1.54E+11 1.46E+11 1.38E+11 cs134 2.87E+13 1.35E+13 6.37E+12 3.00E+12 1.41E+12 6.66E+11 3.13E+11 eu154 1.20E+12 1.00E+12 8.37E+11 6.99E+11 5.83E+11 4.87E+11 4.06E+11


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d y 90 1.19E+10 1.09E+10 1.03E+10 9.75E+09 9.22E+09 8.73E+09 8.26E+09 rh106 6.62E+12 1.33E+12 2.89E+11 6.28E+10 1.36E+10 2.97E+09 6.45E+08 pr144 4.36E+13 5.92E+12 8.09E+11 1.11E+11 1.51E+10 2.06E+09 2.82E+08 eu154 6.44E+08 5.37E+08 4.49E+08 3.74E+08 3.12E+08 2.61E+08 2.18E+08


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d rh106 3.70E+12 7.43E+11 1.61E+11 3.51E+10 7.63E+09 1.66E+09 3.60E+08 pr144 4.12E+11 5.60E+10 7.65E+09 1.04E+09 1.43E+08 1.95E+07 2.66E+06


initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d rh106 6.24E+11 1.25E+11 2.72E+10 5.92E+09 1.29E+09 2.80E+08 6.08E+07



S T U K - Y T O - T R 1 7 0

31



principal photon sources in group 16, photons/secmean energy = 4.2500 mev. nuclides exceeding 1.0E-02 of total group release rate (6.91E-06) at4091.7 dnuclide time after discharge

initial 818.3 d 1636.7 d 2455.0 d 3273.3 d 4091.7 d 4910.0 d ce142 5.77E-06 5.77E-06 5.77E-06 5.77E-06 5.77E-06 5.77E-06 5.77E-06 sm147 4.44E-07 7.46E-07 9.13E-07 1.01E-06 1.06E-06 1.08E-06 1.10E-06





GAMMA SOURCE OF SPENT BWR FUEL APPENDIX D

assessment of the neutron and gamma sources of the spent

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